ML021420362

Part a - Vogtle Electric Generating Plant - 2001 Annual Report

ML021420362
Person / Time
Site:	Vogtle
Issue date:	05/13/2002
From:	Beasley J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCV-1621
Download: ML021420362 (178)

Text

J. Barnie Beasley, Jr., RE.

Vice President Vogtle Project Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway P.O. Box 1295 Birmingham, Alabama 35201 Tel 205.992.7110 Fax 205.992.0403 May 13, 2002 Docket Nos.

SOUTHERNA COMPANY Energy to Serve Your World LCV-1621 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 VOGTLE ELECTRIC GENERATING PLANT 2001 ANNUAL REPORT Gentlemen:

In accordance with the applicable regulatory requirements, Southern Nuclear Operating Company (SNC) hereby submits the 2001 Annual Report of operating information.

The remainder of the 2001 reports not previously submitted is included.

Sincerely, JBB/JLL

Enclosure:

2001 Annual Report A.O0i

U. S. Nuclear Regulatory Commission Page 2 xc:

Southern Nuclear Operating Company Mr. J. T. Gasser (w/o)

Mr. M. Sheibani (w/o)

SNC Document Management Georgia Power Company Mr. M. C. Nichols U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. F. Rinaldi, Project Manager, NRR Mr. J. Zeiler, Senior Resident Inspector, Vogtle State of Georgia Mr. J. L. Setser, DNR American Nuclear Insurers Mr. R. Oliveira

VOGTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 NRC DOCKET NOS. 50-424 AND 50-425 FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 2001 ANNUAL REPORT TABLE OF CONTENTS I.

INTRODUCTION II.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT III.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT IV.

OFFSITE DOSE CALCULATION MANUAL - REVISION 19

VOGTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 2001 ANNUAL REPORT INTRODUCTION The Vogtle Electric Generating Plant Units 1 and 2 are powered by pressurized water reactors; each rated at 3565 megawatts thermal. It is located on the Savannah River in Burke County, Georgia, at a site 34 miles southeast of Augusta.

The Unit 1 operating license was initially received on January 16, 1987, and commercial operation started on May 31, 1987. Unit 1 is in its eleventh fuiel cycle.

Unit 2 received its initial operating license on February 9, 1989, and began commercial operation on May 19, 1989. Unit 2 is operating in its ninth fuel cycle.

VOGTLE ELECTRIC GENERATING PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2001 4Mr SOUTHERNAN COMPANY Energy to Se rve Yoru" World

TABLE OF CONTENTS Section and/or Title Subsection Page List of Figures ii List of Tables ii List of Acronyms iv 1.0 Introduction 1-1 2.0 REMP Description 2-1 3.0 Results Summary 3-1 4.0 Discussion of Results 4-1 4.1 Land Use Census and River Survey 4-6 4.2 Airborne 4-8 4.3 Direct Radiation 4-11 4.4 Milk 4-16 4.5 Vegetation 4-18 4.6 River Water 4-20 4.7 Drinking Water 4-23 4.8 Fish 4-29 4.9 Sediment 4-32 5.0 Interlaboratory Comparison Program (ICP) 5-1 6.0 Conclusions 6-1 i

LIST OF FIGURES Figure Number Title Page Figure 2-1 REMP Stations in the Plant Vicinity 2-10 Figure 2-2 REMP Control Stations for the Plant 2-11 Figure 2-3 REMP Indicator Drinking Water Stations 2-12 Figure 4.2-1 Average Weekly Gross Beta Air Concentration 4-9 Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 4-12 Figure 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 4-13 Figure 4.4-1 Average Annual Cs-137 Concentration in Milk 4-16 Figure 4.5-1 Average Annual Cs-137 Concentration in Vegetation 4-19 Figure 4.6-1 Average Annual H-3 Concentration in River Water 4-21 Figure 4.7-1 Average Monthly Gross Beta Concentration in Raw Drinking Water 4-24 Figure 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water 4-25 Figure 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water 4-27 Figure 4.7-4 Average Annual H-3 Concentration in Finished Drinking Water 4-28 Figure 4.8-1 Average Annual Cs-137 Concentration in Fish 4-30 Figure 4.9-1 Average Annual Be-7 Concentration in Sediment 4-34 Figure 4.9-2 Average Annual Co-58 Concentration in Sediment 4-35 Figure 4.9-3 Average Annual Co-60 Concentration in Sediment 4-36 Figure 4.9-4 Average Annual Cs-137 Concentration in Sediment 4-37 ii

LIST OF TABLES Table Number Title Page Table 2-1 Summary Description of Radiological Environmental Monitoring Program 2-2 Table 2-2 Radiological Environmental Sampling Locations 2-7 Table 3-1 Radiological Environmental Monitoring Program Annual Summary 3-2 Table 4-1 Minimum Detectable Concentrations (MDC) 4-1 Table 4-2 Reporting Levels (RL) 4-2 Table 4-3 Deviations from Radiological Environmental Monitoring Program 4-4 Table 4.1-1 Land Use Census Results 4-6 Table 4.2-1 Average Weekly Gross Beta Air Concentration 4-9 Table 4.3-1 Average Quarterly Exposure from Direct Radiation 4-12 Table 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 4-14 Table 4.4-1 Average Annual Cs-137 Concentration in Milk 4-17 Table 4.5-1 Average Annual Cs-137 Concentration in Vegetation 4-19 Table 4.6-1 Average Annual H-3 Concentration in River Water 4-22 Table 4.7-1 Average Monthly Gross Beta Concentration in Raw Drinking Water 4-24 Table 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water 4-25 Table 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water 4-27 Table 4.7-4 Average Annual H-3 Concentration in Finished Drinking Water 4-28 Table 4.8-1 Average Annual Cs-137 Concentration in Fish 4-30 Table 4.9-1 Average Annual Be-7 Concentration in Sediment 4-34 Table 4.9-2 Average Annual Co-58 Concentration in Sediment 4-35 Table 4.9-3 Average Annual Co-60 Concentration in Sediment 4-36 Table 4.9-4 Average Annual Cs-137 Concentration in Sediment 4-37 Table 4.9-5 Additional Sediment Nuclide Concentrations 4-38 Table 5-1 Interlaboratory Comparison Program Results 5-3 iii

LIST OF ACRONYMS Acronyms presented in alphabetical order.

Acronym Definition ASTM American Society for Testing and Materials CL Confidence Level EL Georgia Power Company Environmental Laboratory EPA Environmental Protection Agency GPC Georgia Power Company ICP Interlaboratory Comparison Program MDC Minimum Detectable Concentration MDD Minimum Detectable Difference MWe MegaWatts Electric NA Not Applicable NDM No Detectable Measurement(s)

NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual Po Preoperation PWR Pressurized Water Reactor REMP Radiological Environmental Monitoring Program RL Reporting Level RM River Mile TLD Thermoluminescent Dosimeter TS Technical Specification VEGP Alvin W. Vogtle Electric Generating Plant iv

1.0 INTRODUCTION

The Radiological Environmental Monitoring Program (REMP) is conducted in accordance with Chapter 4 of the Offsite Dose Calculation Manual (ODCM). The REMP activities for 2001 are reported herein in accordance with Technical Specification (TS) 5.6.2 and ODCM 7.1.

The objectives of the REMP are to:

1) Determine the levels of radiation and the concentrations of radioactivity in the environs and;

2) Assess the radiological impact (if any) to the environment due to the operation of the Alvin W. Vogtle Electric Generating Plant (VEGP).

The assessments include comparisons between results of analyses of samples obtained at locations where radiological levels are not expected to be affected by plant operation (control stations) and at locations where radiological levels are more likely to be affected by plant operation (indicator stations), as well as comparisons between preoperational and operational sample results.

VEGP is owned by Georgia Power Company (GPC), Oglethorpe Power Corporation, the Municipal Electric Authority of Georgia, and the City of Dalton, Georgia. It is located on the southwest side of the Savannah River approximately 23 river miles upstream from the intersection of the Savannah River and U.S.

Highway 301. The site is in the eastern sector of Burke County, Georgia, and across the river from Barnwell County, South Carolina. The VEGP site is directly across the Savannah River from the Department of Energy Savannah River Site.

Unit 1, a Westinghouse Electric Corporation Pressurized Water Reactor (PWR),

with a licensed core thermal power of 3565 MegaWatts (MWt), received its operating license on January 16, 1987 and commercial operation started on May 31, 1987. Unit 2, also a Westinghouse PWR rated for 3565 MWt, received its operating license on February 9, 1987 and began commercial operation on May 19, 1989.

The preoperational stage of the REMP began with initial sample collections in August of 1981. The transition from the pre-operational to the operational stage of the REMP occurred as Unit 1 reached initial criticality on March 9, 1987.

A description of the REMP is provided in Section 2 of this report. Maps showing the sampling stations are keyed to a table which indicates the direction and distance of each station from a point midway between the two reactors. Section 3 provides a summary of the results of the analyses of REMP samples for the year.

The results are discussed, including an assessment of any radiological impacts upon the environment and the results of the land use census and the river survey, in Section 4. The results of the Interlaboratory Comparison Program (ICP) are provided in Section 5. Conclusions are provided in Section 6.

1-1

2.0 REMP DESCRIPTION A summary description of the REMP is provided in Table 2-1.

This table summarizes the program as it meets the requirements outlined in ODCM Table 4

1. It details the sample types to be collected and the analyses to be performed in order to monitor the airborne, direct radiation, waterborne and ingestion pathways, and also delineates the collection and analysis frequencies. In addition, Table 2-1 references the locations of stations as described in ODCM Section 4.2 and in Table 2-2 of this report. The stations are also depicted on maps in Figures 2-1 through 2-3.

REMP samples are collected by Georgia Power Company's (GPC) Environmental Laboratory (EL) personnel. The same lab performs all the laboratory analyses at their headquarters in Smyrna, Georgia.

2-1

TABLE 2-1 (SHEET 1 of 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Number of Sampling and Collection Type and Frequency of Sample Representative Samples Frequency Analysis and Sample Locations

1. Direct Radiation Thirty nine routine Quarterly Gamma dose, quarterly monitoring stations with two or more dosimeters placed as follows:

An inner ring of stations, one in each compass sector in the general area of the site boundary; An outer ring of stations, one in each compass sector at approximately 5 miles from the site; and Special interest areas, such as population centers, nearby recreation areas, and control stations.

2. Airborne Radioiodine and Samples from seven Continuous sampler operation Radioiodine canister: I Particulates locations:

with sample collection weekly, or 131 analysis, weekly.

more frequently if required by Five locations close to dust loading.

Particulate sampler:

the site boundary in Gross beta analysis' different sectors; following filter change and gamma isotopic A community having the analysis 2 of composite highest calculated annual (by location), quarterly.

average ground level D/Q; and

TABLE 2-1 (SHEET 2 of 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Number of Sampling and Collection Type and Frequency of Sample Representative Samples Frequency Analysis and Sample Locations

2. Airborne Radioiodine and A control location near Particulates (cont.)

a population center at a distance of about 14 miles.

3. Waterborne

a. Surface3 One sample upriver.

Composite sample over one Gamma isotopic month period4.

analysis 2, monthly.

Two samples Composite for tritium downriver.

analysis, quarterly.

b..............................

I...........................................................................

I I..

Dr in k in g....

~ e c h ~

I.................

s a.............e

............ r wa............r.........

h b.Drinking Two samples at each of Composite sample of river water 1-13 1 analysis on each the two nearest water near the intake of each water sample when the dose treatment plants that treatment plant over two week calculated for the could be affected by period 4 when 1-131 analysis is consumption of the plant discharges.

required for each sample; water is greater than 1 monthly composite otherwise; and mrem per year.

Two samples at a grab sample of finished water at Composite for gross control location, each water treatment plant every beta and gamma two weeks or monthly, as isotopic analysis2 on appropriate.

raw water, monthly.

Gross beta, gamma isotopic and 1-131 analyses on grab sample of finished water, monthly. Composite for tritium analysis on raw and finished water, quarterly.

aIm...........

c. Sediment from Shoreline One sample from Semiannually Gamma isotopic downriver area with analysis2, semiannually.

existing or potential recreational value.

TABLE 2-1 (SHEET 3 of 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Number of Sampling and Collection Type and Frequency of Sample Representative Samples Frequency Analysis and Sample Locations

c. Sediment from Shoreline One sample from (cont.)

upriver area with existing or potential recreational value.

a. Milk Two samples from Biweekly Gamma isotopic milking animals 6 at analysis 2'7, biweekly.

control locations at a distance of about 10 miles or more.

b...

.Fish..

tl-At t least s"a pone e-.

s am

'f'pie

.of...Semiannually....amma,-isGamma

• ** i......

p any commercially or analysis 2 on edible recreationally portions, semiannually.

important species near the plant discharge.

At least one sample of any commercially or recreationally important species in an area not influenced by plant discharges.

At least one sample of During the spring spawning Gamma isotopic any anadromous season.

analysis2 on edible species near the plant portions, annually.

discharge.

TABLE 2-1 (SHEET 4 of 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Number of Sampling and Collection Type and Frequency of Sample Representative Samples Frequency Analysis and Sample Locations

c. Grass or Leafy Vegetation One sample from two Monthly during growing season.

Gamma isotopic onsite locations near the analysis2' 7, monthly.

site boundary in different sectors.

One sample from a control location at a distance of about 17 miles.

TABLE 2-1 (SHEET 5 of 5)

SUMMARY

DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Notes:

(1)

Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

(2)

Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

(3)

Upriver sample is taken at a distance beyond significant influence of the discharge. Downriver samples are taken beyond but near the mixing zone.

(4)

Composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) to assure obtaining a representative sample.

(5)

The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

(6)

A milking animal is a cow or goat producing milk for human consumption.

(7)

If the gamma isotopic analysis is not sensitive enough to meet the Minimum Detectable Concentration (MDC) for 1-131, a separate analysis for 1-131 may be performed.

TABLE 2-2 (SHEET 1 of 3)

RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Direction1 Distance Sample Type Number Type Location (miles)1 1

Indicator River Bank N

1.1 Direct Rad.

2 Indicator River Bank NNE 0.8 Direct Rad.

3 Indicator Discharge Area NE 0.6 Airborne Rad.

3 Indicator River Bank NE 0.7 Direct Rad 4

Indicator River Bank ENE 0.8 Direct Rad.

5 Indicator River Bank E

1.0 Direct Rad.

6 Indicator Plant Wilson ESE 1.1 Direct Rad.

7 Indicator Simulator SE 1.7 Airborne Rad.

Building Direct Rad.

Vegetation 8

Indicator River Road SSE 1.1 Direct Rad.

9 Indicator River Road S

1.1 Direct Rad.

10 Indicator Met Tower SSW 0.9 Airborne Rad.

10 Indicator River Road SSW 1.1 Direct Rad.

11 Indicator River Road SW 1.2 Direct Rad.

12 Indicator River Road WSW 1.2 Airborne Rad.

Direct Rad.

13 Indicator River Road W

1.3 Direct Rad.

14 Indicator River Road WNW 1.8 Direct Rad.

15 Indicator Hancock NW 1.5 Direct Rad.

Landing Road Vegetation 16 Indicator Hancock NNW 1.4 Airborne Rad.

Landing Road Direct Rad.

17 Other Sav. River Site N

5.4 Direct Rad.

(SRS), River Road 18 Other SRS, D Area NNE 5.0 Direct Rad.

19 Other SRS, Road NE 4.6 Direct Rad.

A.13 20 Other SRS, Road ENE 4.8 Direct Rad.

A. 13.1 21 Other SRS, Road E

5.3 Direct Rad.

A.17 22 Other River Bank ESE 5.2 Direct Rad.

23 Other River Road SE 4.6 Direct Rad.

24 Other Chance Road SSE 4.9 Direct Rad.

25 Other Chance Road S

5.2 Direct Rad.

near Highway 23 26 Other Highway 23 SSW 4.6 Direct Rad.

and Ebenezer Church Road 27 Other Highway 23 SW 4.7 Direct Rad.

opposite Boll Weevil Road 28 Other Thomas Road WSW 5.0 Direct Rad.

2-7

TABLE 2-2 (SHEET 2 of 3)

RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Direction1 Distance Sample Type Number Type Location (miles)1 29 Other Claxton-Lively W

5.1 Direct Rad.

Road 30 Other Nathaniel WNW 5.0 Direct Rad.

Howard Road 31 Other River Road at NW 5.0 Direct Rad.

Allen's Chapel Fork 32 Other River Bank NNW 4.7 Direct Rad.

35 Other Girard SSE 6.6 Airborne Rad.

Direct Rad.

36 Control GPC WSW 13.9 Airborne Rad.

Waynesboro Op.

Direct Rad.

37 Control Substation WSW 16.7 Direct Rad Waynesboro, Vegetation GA 43 Other Employee's Rec.

SW 2.2 Direct Rad.

Center 47 Control Oak Grove SE 10.4 Direct Rad.

Church 48 Control McBean NW 10.2 Direct Rad.

Cemetery 51 Control SGA School S

11.0 Direct Rad.

Sardis, GA 52 Control Oglethorpe SW 10.7 Direct Rad.

Substation; Alexander, GA 80 Control Augusta Water NNW 29.0 Drinking Treatment Plant Water2 81 Control Sav River N

2.5 Fish3 Sediment4 82 Control Sav River (RM NNE 0.8 River Water 151.2) 83 Indicator Sav River (RM ENE 0.8 River Water 150.4)

Sediment4 84 Other Sav River (RM ESE 1.6 River Water 149.5) 85 Indicator Say River ESE 4.3 Fish3 87 Indicator Beaufort-Jasper SE 76 Drinking County Water Water5 Treatment Plant 88 Indicator Cherokee Hill SSE 72 Drinking Water Treatment Water6 Plant, Port Wentworth, Ga 98 Control W.C. Dixon SE 9.8 Milk Dairy 99 Control Boyceland Dairy W

20.9 Milk 2-8

TABLE 2-2 (SHEET 3 of 3)

RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Notes:

(1)

Direction and distance are determined from a point midway between the two reactors.

(2)

The intake for the Augusta Water Treatment Plant is located on the Augusta Canal.

The entrance to the canal is at River Mile (RM) 207 on the Savannah River. The canal effectively parallels the river. The intake to the pumping station is about 4 miles down the canal.

(3)

A 5 mile stretch of the river is generally needed to obtain adequate fish samples.

Samples are normally gathered between RM 153 and 158 for upriver collections and between RM 144 and 149.4 for downriver collections.

(4)

Sediment is collected at locations with existing or potential recreational value.

Because high water, shifting of the river bottom, or other reasons could cause a suitable location for sediment collections to become unavailable or unsuitable, a stretch of the river between RM 148.5 and 150.5 was designated for downriver collections while a stretch between RM 153 and 154 was designated for upriver collections. In practice, collections are normally made at RM 150.2 for downriver collections and RM 153.3 for upriver collections.

(5)

The intake for the Beaufort-Jasper County Water Treatment Plant is located at the end of a canal that begins at RM 39.3 on the Savannah River. This intake is about 16 miles by line of sight down the canal from its beginning on the Savannah River.

(6)

The intake for the Cherokee Hill Water Treatment Plant is located on Abercorn Creek which is about one and a quarter creek miles from its mouth on the Savannah River at RM 29.

2-9

Radiological Environmental Sampling Locations Indicator Contrd Additional REMP Stations in the I T A

A A

Plant Vicinity Other 0

6 TL & Other

© Figure 2-1 2-10 CIO I

Radiological Environmental Sampling Locations Indicator Control Additional REMP Control Stations TLD A

A A

for the Plant Other 0

0 0

TLD & Other 0

0 0

Figure 2-2 2-11

Savanne A"

NO4

~

Radiological Environmental Sampling Locations Indicator Contol Addtnal REMP Indicator Drinking TLD A

A A

Water Stations Other S

0 0

UD &other 0iFigure2-3 2-12 c002'

3.0 RESULTS

SUMMARY

In accordance with ODCM 7.1.2.1, the summarized and tabulated results for all of the regular samples collected for the year at the designated indicator and control stations are presented in Table 3-1. The format of Table 3-1 is similar to Table 3 of the Nuclear Regulatory Commission (NRC) Branch Technical Position, "An Acceptable Radiological Environmental Monitoring Program", Revision 1, November 1979. Results for samples collected at locations other than indicator or control stations are discussed in Section 4 under the particular sample type.

As indicated in ODCM 7.1.2.1, the results for naturally occurring radionuclides that are also found in plant effluents must be reported along with man-made radionuclides. The radionuclide Be-7 which occurs abundantly in nature is found in some years in the plant's liquid and gaseous effluent. No other naturally occurring radionuclides are found in the plant's effluent releases. Therefore, the only radionuclides of interest in the REMP samples are the man-made radionuclides and Be-7, when it is detected in the effluent.

3-1

TABLE 3-1 (SHEET 1 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

Airborne Gross Beta 10 22.4 Station 16 22.9 22.0 Particulates 318 7-48 Hancock 7-39 10-44 (fCi/m3)

(265/265)

Landing Road (53/53)

(53/53) 1.4 miles NNW Gamma Isotopic 24 Be-7 24 69.2 Station 12 77.0 71.9 47-101 River Road 64-89 60-94

.....(20/20)..........

1.2 m iles W SW (4 )(

/ )

Cs-134 50 NDM (c)

NDM NDM Cs-137 60 NDM NDM NDM Airborne 1-131 70 NDM NDM NDM Radioiodine 318 (fCi/m3)

Direct Gamma NA (d) 12.9 Station 20 17.1 13.0 Radiation Dose 9.2-17.7 SRS Road 14.8-18.2 10.9-17.0 (mR/91 days) 87 (63/63)

A. 13.1 (4/4)

(24/24) 4.8 miles ENE

TABLE 3-1 (SHEET 2 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

Milk (pCi/1)

Gamma Isotopic 52 Be-7 124 NA NDM NDM Cs-134 15 NA NDM NDM Cs-137 18 NA NDM NDM Ba-140 60 NA NDM NDM La-140 15 NA NDM NDM I-131 1

NA NDM NDM 52 Vegetation Gamma (pCi/kg-wet)

Isotopic 36 Be-7 729 1616.4 Station 07 1659.0 1260.1 235-4039 Simulator Bldg.

143-3704 194-3779 (2 / 4 )

.1.7.m ile sSS EE..............................(1 2 /1 2).................................

(142 /1 2).................7....E...................!2 1...............................1 2

-- 131 60 NDM NDM NDM Cs-134 60 NDM NDM NDM Cs-137 80 NDM NDM NDM

TABLE 3-1 (SHEET 3 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

River Water Gamma (pCi/1)

Isotopic 24 Be-7 124 NDM NDM NDM Mn-54 15 NDM NDM NDM Fe-59 30 NDM NDM NDM Co-58 15 NDM NDM NDM Co-60 15 NDM NDM NDM Zn-65 30 NDM NDM NDM Zr-95 30 NDM NDM NDM Nb-95 15 NDM NDM NDM 1-131 15 NDM NDM NDM Cs-134 15 NDM NDM NDM Cs-137 18 NDM NDM NDM Ba-140 60 NDM NDM NDM La-140 15 NDM NDM NDM Tritium 3000 2101 Station 83 2101 743 8

1602-3200 Downriver 1602-3200 676-798 (4/4) 0.8 miles (4/4)

(4/4) 0.4 RM

TABLE 3-1 (SHEET 4 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

Water Near Gross Beta 4

3.21 Station 87 3.49 2.94 Intakes to 36 0.51-7.45 Beaufort, SC 1.11-7.45 0.58-7.55 Water (24/24)

Downriver (12/12)

(12/12)

Treatment 76 miles Plants (pCi/1)

(112 RM)

Gamma Isotopic 36 Be-7 124 (e)

NDM NDM NDM Mn-54 15 NDM NDM NDM Fe-59 30 NDM NDM NDM Co-58 15 NDM NDM NDM Co-60 15 NDM NDM NDM Zn-65 30 NDM NDM NDM Zr-95 30 NDM NDM NDM Nb-95 15 NDM NDM NDM 1-131(f) 15 NDM NDM NDM Cs-134 15 NDM NDM NDM Cs-137 18 NDM NDM NDM Ba-140 60 NDM NDM NDM La-140 15 NDM NDM NDM T.im3 0 08 9S a.o 794 Triium300 89 tation 87 94525 12 452-1300 Beaufort 794-1300 525-525 (8/8)

Downriver (4/4)

(1/4) 76 miles (112 RM)

TABLE 3-1 (SHEET 5 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

Finished Water Gross Beta 4

2.67 Station 88 2.71 2.00 at Water 36 0.45-6.19 Port Wentworth 0.45-4.75 1.06-4.08 Treatment (24/24)

Downriver (12/12)

(12/12)

Plants (pCi/1) 72miles (122 RM)

Gamma Isotopic 36 Be-7 124 (e)

NDM NDM NDM Mn-54 15 NDM NDM NDM Fe-59 30 NDM NDM NDM Co-58 15 NDM NDM NDM Co-60 15 NDM NDM NDM Zn-65 30 NDM NDM NDM Zr-95 30 NDM NDM NDM Nb-95 15 NDM NDM NDM 1-131 1

NDM NDM NDM Cs-134 15 NDM NDM NDM Cs-137 18 NDM NDM NDM Ba-140 60 NDM NDM NDM La-140 15 NDM NDM NDM Tritium 2000 1037 Station 88 1068 516 12 613-1780 Port Wentworth 613-1780 516-516 (8/8)

Downriver (4/4)

(1/4) 72 miles S(122 RM)

TABLE 3-1 (SHEET 6 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement) Performed (Fraction)

& Direction Range (Fraction)

Anadromous Gamma Fish Isotopic (pCi/kg-wet) 1 Be-7 655 (e)

NDM NDM NA Mn-54 130 NDM NDM NA Fe-59 260 NDM NDM NA Co-58 130 NDM NDM NA Co-60 130 NDM NDM NA Zn-65 260 NDM NDM NA Cs-134 130 NDM NDM NA Cs-137 150 NDM NDM NA Fish Gamma (pCi/kg-wet)

Isotopic 8

Be-7 655 (e)

NDM NDM NDM Mn-54 130 NDM NDM NDM Fe-59 260 NDM NDM NDM Co-58 130 NDM NDM NDM Co-60 130 NDM NDM NDM Zn-65 260 NDM NDM NDM Cs-134 130 NDM NDM NDM Cs-137 150 47.6 Station 85 47.6 39.1 41.4-53.8 Downriver 41.4-53.8 24.3-53.8 (2/3) 4.3 miles (2/3)

(2/4)

TABLE 3-1 (SHEET 7 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Detectable Locations Annual Mean Locations Mean Sampled Number of Concentration Mean (b),

(b), Range (Unit of Analyses (MDC) (a)

Range Name Distance Mean (b),

(Fraction)

Measurement)

Performed (Fraction)

& Direction Range (Fraction)

Sediment Gamma (pCi/kg-dry)

Isotopic 4

Be-7 655(e) 1697 Station 81 2614 2614 133-3262 Upriver 2614-2614 2614-2614

.... o

6.P..................

.. O e )....

. P..........

I IP Cs-134 10NDM NDM NDM Cs-137 180 252.0 Station 83 252.0 117.6 252-252 Downriver 252-252 33.2-202.0 (1/2) 0.8 miles (1/2)

(2/2)

TABLE 3-1 (SHEET 8 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL

SUMMARY

Vogtle Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Notes:

a. The MDC is defined in ODCM 10.1. Except as noted otherwise, the values listed in this column are the detection capabilities required by ODCM Table 4-3. The values listed in this column are a priori (before the fact) MDCs. In practice, the a posteriori (after the fact) MDCs are generally lower than the values listed. Any a posteriori MDC greater than the value listed in this column is discussed in Section 4.

b. Mean and range are based upon detectable measurements only. The fraction of all measurements at a specified location that are detectable is placed in parenthesis.

c. No Detectable Measurement(s).

d. Not Applicable.

e. The EL has determined that this value may be routinely attained under normal conditions. No value is provided in ODCM Table 4-3.

f. Item 3 of ODCM Table 4-1 implies that an 1-131 analysis is not required to be performed on water samples when the dose calculated from the consumption of water is less then 1 mrem per year. However, 1-131 analyses have been performed on the finished drinking water samples.

4.0 DISCUSSION OF RESULTS Included in this section are evaluations of the laboratory results for the various sample types. Comparisons were made between the difference in mean values for pairs of station groups (e.g., indicator and control stations) and the calculated Minimum Detectable Difference (MDD) between these pairs at the 99%

Confidence Level (CL). The MDD was determined using the standard Student's t test. A difference in the mean values that was less than the MDD was considered to be statistically indiscernible.

The 2001 results were compared with past results, including those obtained during preoperation.

As appropriate, results were compared with their Minimum Detectable Concentrations (MDC) and Reporting Levels (RL) which are listed in Tables 4-1 and 4-2 of this report, respectively. The required MDCs were achieved during laboratory sample analysis. Any anomalous results are explained within this report.

Results of interest are graphed to show historical trends. The data points are tabulated and included in this report. The points plotted and provided in the tables represent mean values of only detectable results. Periods for which no detectable measurements (NDM) were observed or periods for which values were not applicable (e.g., milk indicator, etc.) are plotted as and listed in the tables as 0's.

Table 4-1 Minimum Detectable Concentrations (MDC)

Analysis Water Airborne Fish Milk Grass or Sediment (pCi/1)

Particulate (pCi/kg-(pCi/I)

Leafy (pCi/kg) or Gases wet)

Vegetation (fCi/m3)

(pCi/kg wet)

Gross Beta 4

10 H-3 2000 (a)

Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 1 (b) 70 1

60 Cs-134 15 50 130 15 60 150 Cs-137 18 60 150 18 80 180 Ba-140 60 60 La-140 15 15 4-1 (a) If no drinking water pathway exists, a value of 3000 pCi/l may be used.

(b) If no drinking water pathway exists, a value of 15 pCi/l may be used.

Table 4-2 Reporting Levels (RL)

Analysis Water Airborne Fish Milk (pCi/l)

Grass or (pCi/1)

Particulate (pCi/kg-wet)

Leafy or Gases Vegetation (fCi/m3)

(pCi/kg-wet)

H-3 20,000 (a)

Mn-54 1000 30,000 Fe-59 400 10,000 Co-58 1000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-95 400 Nb-95 700 1-131 2(b) 900 3

100 Cs-134 30 10,000 1000 60 1000 Cs-137 50 20,000 2000 70 2000 Ba-140 200

_300 La-140 100 1

400 (a) This is the 40 CFR 141 value for drinking water samples.

pathway exists, a value of 30,000 may be used.

If no drinking water (b) If no drinking water pathway exists, a value of 20 pCi/1 may be used.

Atmospheric nuclear weapons tests from the mid 1940s through 1980 distributed man-made nuclides around the world. The most recent atmospheric tests in the 1970s and in 1980 had a significant impact upon the radiological concentrations found in the environment prior to and during preoperation, and the earlier years of operation. Some long lived radionuclides, such as Cs-137, continue to have some impact. A significant component of the Cs-137 which has often been found in various samples over the years (and continues to be found) is attributed to the nuclear weapons tests.

Data in this section has been modified to remove any obvious non-plant short term impacts. The specific short term impact data that has been removed includes: the nuclear atmospheric weapon test in the fall of 1980; abnormal releases from the Savannah River Site (SRS) during 1987 and 1991; and the Chernobyl incident in the spring of 1986.

In accordance with ODCM 4.1.1.2.1, deviations from the required sampling schedule are permitted, if samples are unobtainable due to hazardous conditions, unavailability, inclement weather, equipment malfunction or other just reasons.

Deviations from conducting the REMP as described in Table 2-1 are summarized in Table 4-3 along with their causes and resolutions. As discussed in Section 4.3, during 2001 only one deviation resulted in loss of data. During quarter 2, Station 09 was vandalized and the TLDs were damaged beyond recovery.

4-2

All results were tested for conformance with Chauvenet's criterion (G. D. Chase and J. L. Rabinowitz, Principles of Radioisotope Methodology, Burgess Publishing Company, 1962, pages 87-90) to identify values which differed from the mean of a set by a statistically significant amount. Identified outliers were investigated to determine the reason(s) for the difference.

If equipment malfunction or other valid physical reasons were identified as causing the variation, the anomalous result was excluded from the data set as non representative.

No data were excluded exclusively for failing Chauvenet's criterion.

Data exclusions are discussed in this section under the appropriate sample type.

4-3

TABLE 4-3 (SHEET 1 of 2)

DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLES 03/20/01 -

Simulator Airborne particulates and Cause determined 04-03-01 to Replaced clock on 04/03/01.

03/27/01 Air Filter and radioiodine monitoring were not be malfunctioning station clock.

Air Cartridge performed for 23.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

Station 03/27/01 -

Simulator Airborne particulates and Cause determined 04-03-01 to Replaced clock on 04/03/01.

04/03/01 Air Filter and radioiodine monitoring were not be malfunctioning station clock.

Air Cartridge performed for 41.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Station 05/29/01 -

Hancock Air Airborne particulates and Power outage due to tree limb Line repaired, power was 06/05/01 Filter and Air radioiodine monitoring were not on power line caused by severe restored and air sampler Cartridge Station performed on 06/04/01 for 2.4 weather.

returned to service.

hours.

05/29/01 -

Met Tower Air Airborne particulates and Power outage due to tree limb Line repaired, power was 06/05/01 Filter and Air radioiodine monitoring were not on power line caused by severe restored and air sampler Cartridge Station performed on 06/04/01 for 5.1 weather.

returned to service.

hours.

I II

TABLE 4-3 (SHEET 2 of 2)

DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLES 05/29/01 -

Girard Air Filter Airborne particulates and Power outage due to tree limb Line repaired, power was 06/05/01 and Air Cartridge radioiodine monitoring were not on power line caused by severe restored and air sampler returned Stations performed on 06/03 and 06/04/01 weather.

to service.

for 9.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

06/19/01-Girard Air Filter Airborne radioiodine sample Upon change-out, charcoal Staff was briefed on adequate 06/26/01 and Air Cartridge collection was not representative, cartridge was found to be care during charcoal cartridge Stations damaged.

handling and installation.

06/19/01 -

Waynesboro Airborne particulates and Power outage due to faulty Cable was repaired and air 06/26/01 Air Filter and Air radioiodine monitoring were not electrical cable.

sampler returned to service.

Cartridge Station performed for 39.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

06/26/01-Waynesboro Airborne particulates and Power outage due to faulty Cable was repaired and air 07/03/01 Air Filter and Air radioiodine monitoring were not electrical cable.

sampler returned to service Cartridge Station performed for about 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

2nd Quarter 01 Milk (ICP No milk sample was obtained Administrative error in that steps Arrangements made with vendor Sample) from the vendor that supplies were not taken to assure that the supplying ICP samples that all ICP samples for analysis milk sample would be supplied, appropriate samples will be furnished to EL for analyses.

2nd Quarter 01 TLD 09 TLD data unavailable TLD was vandalized TLDs replaced.

4th Quarter 01 TLD 21 TLD did not remain in proper TLD at this location was found TLDs replaced.

location for the entire quarter.

lying on ground on the change out date.

4.1 Land Use Census and River Survey In accordance with ODCM 4.1.2, a land use census was conducted on December 11, 2001 to determine the locations of the nearest permanent residence, milk animal, and garden of greater than 500 square feet producing broad leaf vegetation, in each of the 16 compass sectors within a distance of 5 miles; the locations of the nearest beef cattle in each sector were also determined. A milk animal is a cow or goat producing milk for human consumption. Land within SRS was excluded from the census. The census results are tabulated in Table 4.1-1.

Table 4.1-1 LAND USE CENSUS RESULTS Distance in Miles to the Nearest Location in Each Sector SECTOR RESIDENCE MILK BEEF GARDEN ANIMAL CATTLE N

None None None None NNE None None None None NE None None None None ENE None None None None E

None None None None ESE 4.2 None None None SE 4.4 None 5.0 4.8 SSE 4.6 None 4.6 None S

4.4 None None None SSW 4.7 None 4.5 None SW 2.7 None 2.7 4.8 WSW 1.2 None 4.5 None W

3.7 None 4.4 None WNW 1.8 None 2.4 None NW 1.6 None 1.9 2.5 NNW 1.5 None None None ODCM 4.1.2.2.1 requires a new controlling receptor to be identified, if the land use census identifies a location that yields a calculated receptor dose greater than the one in current use. It was determined that no change in the controlling receptor was required in 2001.

ODCM 4.1.2.2.2 requires that whenever the land use census identifies a location which yields a calculated dose (via the same ingestion pathway) 20% greater than that of a current indicator station, the new location must become a REMP station (if samples are available). None of the identified locations yielded a calculated dose 20% greater than that for any of the current indicator stations. No milk animals were identified within five miles of the plant.

4-6

A survey of the Savannah River downstream of the plant for approximately 100 miles was conducted on September 18, 2001 to identify any withdrawal of water from the river for drinking or irrigation purposes. No such usage was identified.

These results were corroborated by checking with the Georgia Department of Natural Resources and the South Carolina Department of Health and Environmental Control. Each of these agencies confirmed that no water withdrawal permits for drinking or irrigation purposes had been issued for this stretch of the Savannah River. The two water treatment plants used as indicator stations for drinking water are located farther downriver.

4-7

4.2 Airborne As specified in Table 2-1 and shown in Figures 2-1 through 2-3, airborne particulate filters and charcoal canisters are collected weekly at 5 indicator stations (Stations 3, 7, 10, 12 and 16) which encircle the plant at the site periphery, at a nearby community station (Station 35) approximately 7 miles from the plant, and at a control station (Station 36) which is approximately 14 miles from the plant. At each location, air is continuously drawn through a glass fiber filter to retain airborne particulate and an activated charcoal canister is placed in series with the filter to adsorb radioiodine.

Each particulate filter is counted for gross beta activity. A quarterly gamma isotopic analysis is performed on a composite of the air particulate filters for each station.

Each charcoal canister is analyzed for 1-131.

As provided in Table 3-1, the 2001 annual average weekly gross beta activity was 22.4 fCi/m 3 for the indicator stations. Although this concentration is slightly higher than in previous years of operation, it was only 0.4 fCi/m 3 greater than the control stations' average for the year. This difference is not statistically discernible, since it is less than the calculated MDD of 2.5 fCi/m 3.

The 2001 annual average weekly gross beta activity at the Girard community station was 22.7 fCi/m 3 which was 0.7 fCi/m 3 greater than the control station's average.

This difference is not statistically discernible since it is less than the calculated MDD of 3.3 fCi/m3.

The historical trending of the average weekly gross beta air concentrations for each year of operation and the preoperational period (September, 1981 to January, 1987) at the indicator, control and community stations is plotted in Figure 4.2-1 and listed in Table 4.2-1.

In general, there is close agreement between the results for the indicator, control and community stations.

This close agreement supports the position that the plant is not contributing significantly to the gross beta concentrations in air.

4-8

Figure 4.2-1 Table 4.2-1 Average Weekly Gross Beta Air Concentration Period Indicator (fCi/m3)

Control Conmmunity (fCi/m3)

(fCi/mS)

Pre-op 22.9 22.1 21.9 1987 26.3 23.6 22.3 1988 24.7 23.7 22.8 1989 19.1 18.2 18.8 1990 19.6 19.4 18.8 1991 19.3 19.2 18.6 1992 18.7 19.3 18.0 1993 21.2 21.4 20.3 1994 20.1 20.3 19.8 1995 21.1 20.7 20.7 1996 23.3 21.0 20.0 1997 20.6 20.6 19.0 1998 22.7 22.4 20.9 1999 22.5 21.9 22.2 2000 24.5 21.5 21.1 2001 22.4 22.0 22.7 4-9 Average Weekly Gross Beta Air Concentration 30 25 E S20 10 0

0 Po 87 88 89 90 91 92 93 94 95 96 91 98 99 00 01 Year

-I--Indicator -U-Control A

-Community D

cor-

During 2001, no man-made radionuclides were detected from the gamma isotopic analysis of the quarterly composites of the air particulate filters. In 1987, Cs-137 was found in one indicator composite at a concentration of 1.7 fCi/m 3. During preoperation, Cs-137 was found in approximately 12% of the indicator composites and 14% of the control composites with average concentrations of 1.7 and 1.0 fCi/m 3, respectively. The MDC for airborne Cs-137 is 60 fCi/m 3. Also, during preoperations, Cs-134 was found in about 8% of the indicator composites at an average concentration of 1.2 fCi/m3. The MDC for Cs-134 is 50 fCi/m3.

The naturally occurring radionuclide Be-7 is typically detected in all indicator and control station gamma isotopic analysis of the quarterly composites of the air particulate filters. Be-7 was identified in plant gaseous effluents in 2001, therefore it is of interest in the REMP samples monitoring the gaseous pathways.

The average Be-7 concentration at the indicator stations was 69.2 fCi/m 3, which is 2.7 fCi/m less than the average concentration found at the control station.

This difference is not statistically discernible because it is less than the calculated MDD of 19.5 fCi/m3. Be-7 has been detected in gaseous effluents eight of the fifteen years of plant operation.

However, there was not a statistically discernible difference between the indicator and control station Be-7 concentrations in air samples in any of the years.

Airborne 1-131 was not detected in any sample during 2001. During preoperation, positive results were obtained only during the Chernobyl incident when concentrations as high as 182 fCi/m were observed.

The MDC and RL for airborne 1-131 are 70 and 900 fCi/m 3, respectively.

Table 4-3 lists REMP deviations that occurred in 2001. None of the deviations listed in Table 4-3 required data to be excluded from the calculation of the mean detectable air sample values.

4-10

4.3 Direct Radiation Direct (external) radiation is measured with thermoluminescent dosimeters (TLDs). Two Panasonic UD-814 TLD badges are placed at each station. Each badge contains three phosphors composed of calcium sulfate crystals (with thulium impurity). The gamma dose at each station is based upon the average readings of the phosphors from the two badges. The badges for each station are placed in thin plastic bags for protection from moisture while in the field.

The badges are nominally exposed for periods of a quarter of a year (91 days). An inspection is performed near mid-quarter to assure that all badges are on-station and to replace any missing or damaged badges.

Two TLD stations are established in each of the 16 compass sectors, to form 2 concentric rings. The inner ring (Stations 1 through 16) is located near the plant perimeter as shown in Figure 2-1 and the outer ring (Stations 17 through 32) is located at a distance of approximately 5 miles from the plant as shown in Figure 2

2. The 16 stations forming the inner ring are designated as the indicator stations.

The two ring configuration of stations was established in accordance with NRC Branch Technical Position "An Acceptable Radiological Environmental Monitoring Program", Revision 1, November 1979.

The 6 control stations (Stations 36, 37, 47, 48, 51 and 52) are located at distances greater than 10 miles from the plant as shown in Figure 2-2. Monitored special interest areas consist of the following: Station 35 at the town of Girard, and Station 43 at the employee recreational area.

As provided in Table 3-1 the average quarterly exposure measured at the indicator stations was 12.9 mR with a range of 9.2 to 17.7 mR. This average was 0.1 mR less than the average quarterly exposure measured at the control stations. This difference is not statistically discernible since it is less than the MDD of 1.1 mR.

Over the operational history of the site, the annual average quarterly exposures shows a variation of no more than 0.7 mR difference between the indicator and control stations. The overall average quarterly exposure for the control stations during preoperation was 1.2 mR greater than that for the indicator stations.

The quarterly exposures acquired at the outer ring stations during 2001 ranged from 10.3 to 17.1 mR with an average of 12.9 mR which was 0.1 mR less than that for the control stations. However, this difference is not discernible since it is less than the MDD of 1.1 mR. For the entire period of operation, the annual average quarterly exposures at the outer ring stations vary by no more than 1.2 mR from those at the control stations. The overall average quarterly exposure for the outer ring stations during preoperation was 1.8 mR less than that for the control stations.

The historical trending of the average quarterly exposures for the indicator inner ring, outer ring, and the control stations are plotted in Figure 4.3-1 and listed in Table 4.3-1. The decrease between 1991 and 1992 values is attributed to a change in TLDs from Teledyne to Panasonic.

It should be noted however that the differences between indicator and control and outer ring values did not change.

The close agreement between the station groups supports the position that the plant is not contributing significantly to direct radiation in the environment.

4-11

Figure 4.3-1 Table 4.3-1 Average Quarterly Exposure from Direct Radiation Period Indicator Control Outer Ring (mR) m)

(mR)

Pre-op 15.3 16.5 14.7 1987 17.6 17.9 16.7 1988 16.8 16.1 16.0 1989 17.9 18.4 17.2 1990 16.9 16.6 16.3 1991 16.9 17.1 16.7 1992 12.3 12.5 12.1 1993 12.4 12.4 12.1 1994 12.3 12.1 11.9 1995 12.0 12.5 12.3 1996 12.3 12.2 12.3 1997 13.0 13.0 13.1 1998 12.3 12.7 12.4 1999 13.6 13.5 13.4 2000 13.5 13.6 13.5 2001 1219 13.0 12.9 4-12 co5 Average Quarterly Exposure from Direct Radiation 20 18 14 12 10 2

0 Po 87 88 80 9fl 91 92 93 94 95 96 97 98 99 00 01 Year I

-indicator

--U-Control Outer Ring

The historical trending of the average quarterly exposures at the special interest areas for the same periods are provided in Figure 4.3-2 and listed in Table 4.3-2.

These exposures are within the range of those acquired at the other stations. They too, show that the plant is not contributing significantly to direct radiation at the special interest areas.

Figure 4.3-2 4-13 COG(

Average Quarterly Exposure from Direct Radiation at Special Interest Areas 25 20

£15 E

o 10 5

Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 Year

[-Htg Cabin (Sta 33) -U--Girard (Sta 35)

- Rec Center (Sta 43)

Table 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas Period Station 33 Station 35 Station 43 Pre-op 16.6 15.1 15.3 1987 21.3 18.5 15.2 1988 19.7 18.1 14.8 1989 21.2 18.7 17.4 1990 16.8 18.9 16.2 1991 17.3 19.6 17.0 1992 12.8 13.5 12.0 1993 12.9 13.3 12.1 1994 12.6 13.6 12.0 1995 13.3 13.5 12.3 1996 13.0 13.6 12.1 1997 13.8 14.4 12.7 1998 13.5 13.7 12.5 1999 NA 14.5 12.7 2000 NA 14.8 13.1 2001 NA 14.0 12.6 The hunting cabin activities at Station 33 have been discontinued and, consequently, this location is no longer considered as an area of special interest.

Monitoring at this location was discontinued at the end of 1998.

There were two deviations from the REMP pertaining to measuring quarterly gamma doses during 2001.

These deviations are listed in Table 4-3.

One deviation occurred during the 2nd quarter when the dosimeter at Station 09 was discovered at quarter's end to have been vandalized.

Therefore, no data were available for this station. The second deviation occurred at Station 21 during the 4th quarter. At collection, this dosimeter was found on the ground, apparently having fallen from its on-station position sometime during the quarter. Therefore, the TLD was not in proper position for the entire quarter.

The resulting data passed Chauvenet's criterion and was not excluded from the data set.

The standard deviation for the quarterly result for each badge was subjected to a self imposed limit of 1.4. This limit is based upon the standard deviations obtained with the Panasonic UD-814 badges during 1992 and is calculated using a method developed by the American Society of Testing and Materials (ASTM Special Technical Publication 15D, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976).

The limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater than 1.4 are excluded since the high standard deviation is interpreted as an indication of unacceptable variation in TLD response.

4-14

The readings for the following badges were deemed unacceptable since the standard deviation for each badge was greater than the self-imposed limit of 1.4:

First Quarter:

None Second Quarter:

14A and 32B Third Quarter:

None Fourth Quarter:

0IB, 15A and 21B However, for these cases when only one badge exceeded a standard deviation of 1.4, the companion badges were available and were used for determining the quarterly doses.

The badges exceeding the self-imposed limit were visually inspected under a microscope and the glow curve and test results for the anneal data and the element correction factors were reviewed. No reason was evident for the high standard deviation.

4-15

4.4 Milk In accordance with Tables 2-1 and 2-2, milk samples are collected biweekly from two control locations, the W. C. Dixon Dairy (Station 98) and the Boyceland Dairy (Station 99). Gamma isotopic and 1-131 analyses are performed on each sample.

No indicator station (a location within 5 miles of the plant) for milk has been available since April 1986. As discussed in Section 4.1, no milk animal was found during the 2001 land use census.

No man-made radionuclide or Be-7 was identified during the gamma isotopic analysis of the milk samples in 2001. The MDC and RL for Cs-137 in milk are 18 and 70 pCi/l, respectively. During preoperation and each year of operation through 1991, Cs-137 was found in 2 to 6% of the samples at concentrations ranging from 5 to 27 pCi/1. During preoperation, Cs-134 was detected in one sample and in the first year of operation, Zn-65 was detected in one sample. Figure 4.4-1 and Table 4.4-1 provide the historical trending of the Cs-137 concentration in milk.

Figure 4.4-1 4-16

"-7J Average Annual Cs-137 Concentration in Milk 20 12 1 6 14 Yea I --

ndicator

-U-Control

-MDC

Table 4.4-1 Average Annual Cs-137 Concentration in Milk Year Indicator Control

__(PCi/l)

I(PCi/I)

Pre-op 18.5 18 1987 0

10.4 1988 0

6.9 1989 0

7 1990 0

17 1991 0

14.2 1992 0

0 1993 0

0 1994 0

0 1995 0

0 1996 0

0 1997 0

0 1998 0

0 1999 0

0 2000 0

0 2001 0

0 During 2001, 1-131 was not detected in any of the milk samples. Since operations began in 1987, 1-131 may have been detected in one sample in 1996 and two during 1990; however, its presence in these cases was questionable, due to large counting uncertainties.

During preoperation, positive 1-131 results were found only during the Chernobyl incident with concentrations ranging from 0.53 to 5.07 pCi/1. The MDC and RL for 1-131 in milk are 1 and 3 pCi/l, respectively.

4-17

4.5 Vegetation In accordance with Tables 2-1 and 2-2, grass samples are collected monthly at two indicator locations onsite near the site boundary (Stations 7 and 15) and at one control station located about 17 miles from the plant (WSW - Station 37). Gamma isotopic analyses are performed on the samples.

During 2001 no man-made radionuclides were detected in grass samples collected at any indicator or control station.

Since Cs-137 is sometimes detected in environmental samples, as a result of atmospheric weapons testing and the Chernobyl incident, the historical trending of the average concentration of Cs-137 at the indicator and control stations is provided in Figure 4.5-1 and listed in Table 4.5-1. No trend is recognized in these data. The MDC and RL for Cs-137 in vegetation samples are 80 and 2000 pCi/kg wet, respectively.

Cs-137 is the only manmade radionuclide that has been identified in vegetation samples during the operational history of the plant. During preoperation, Cs-137 was found in approximately 60% of the samples from indicator stations and in approximately 20% of the samples from the control station. These percentages have generally decreased during operation.

The naturally occurring radionuclide Be-7 is typically detected in indicator and control station vegetation samples.

Be-7 was detected in gaseous effluents in 2001, therefore it is of interest in the REMP samples monitoring the gaseous pathways. The average Be-7 concentration at the indicator stations was 1616 pCi/kg, which is 356 pCi/kg greater than the average concentration found at the control station. This difference is not statistically discernible because it is less than the calculated MDD of 929 pCi/kg. Be-7 has been detected in gaseous effluents eight of the fifteen years of plant operation and is therefore of interest in the REMP program. However, the levels of Be-7 found in the REMP make no significant contribution to dose.

In May and June of 1986 during preoperation, as a consequence of the Chernobyl incident, 1-131 was found in nearly all the samples collected for a period of several weeks in the range of 200 to 500 pCi/kg-wet. The MDC and RL for 1-131 in vegetation are 60 and 100 pCi/kg-wet, respectively. Also during this time period, Co-60 was found in one of the samples at a concentration of 62.5 pCi/kg-wet.

There is no specified MDC or RL for Co-60 in vegetation.

4-18

Figure 4.5-1 Table 4.5-1 Average Annual Cs-137 Concentration in Vegetation Year Indicator Control (pC/kg-wet)

(pCi/kg-wet)

Pre-op 54.6 43.7 1987 24.4 61.5 1988 38.7 0.0 1989 9.7 0.0 1990 30.0 102.0 1991 35.3 62.4 1992 38.1 144.0 1993 46.4 34.1 1994 20.7 57.4 1995 57.8 179.0 1996 0

0 1997 0

32.6 1998 0

50.1 1999 37.2 0

2000 36.6 0

2001 0

0 4-19 Co90 Average Annual Cs-i 37 Concentration in Vegetation 200 ISO

~160

,140 L) 120 C100

.2 20 c~ 0 Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 Year -Indicator

-U*-Control

-MDC

4.6 River Water Surface water from the Savannah River is obtained at three locations using automatic samplers. Small quantities are drawn at intervals not exceeding a few hours.

The samples drawn are collected monthly; quarterly composites are produced from the monthly collections.

The collection points consist of a control location (Station 82) which is located about 0.4 miles upriver of the plant intake structure, an indicator location (Station

83) which is located about 0.4 miles downriver of the plant discharge structure, and a special location (Station 84) which is located approximately 1.3 miles downriver of the plant discharge structure.

A statistically significant increase in the concentrations found in samples collected at the indicator station compared to those collected at the control station could be indicative of plant releases.

Concentrations found at the special station are more likely to represent the activity in the river as a whole, which might include plant releases combined with those from other sources along the river.

A gamma isotopic analysis is conducted on each monthly sample.

As in all previous years, there were no gamma emitting radionuclides of interest detected in the 2001 river water samples.

Each quarterly composite is analyzed for tritium. As indicated in Table 3-1, the average concentration found at the indicator station was 1358 pCi/1 greater than that found at the control station. This difference is not statistically discernible since it is less than the calculated MDD of 1696 pCi/l. At the special station, the results ranged from 881 pCi/1 to 1050 pCi/1 with an average of 931 pCi/1. The MDC for tritium in river water is 3000 pCi/1 and the RL is 30,000 pCi/I.

The historical trending of the average tritium concentrations found at the special, indicator and control stations along with the MDC for tritium is plotted on Figure 4.6-1. The data for the plot is listed in Table 4.6-1. Also included in the table are data from the calculated difference between the indicator and control stations; the MDD between the indicator and control stations; and the total liquid releases of tritium from the plant.

There does not appear to be any good correlation between the plant tritium releases and the differential amount found in the river between the indicator and control stations. The average concentration at the indicator station and its increase over that at the control station for 2001 are about the same as in previous years of operation.

In the first two years of operation, the tritium concentration at the special station was somewhat greater than that at the indicator station. Whereas in recent years, the level at the special station has generally become less than the level at the indicator station.

The annual downriver survey of the Savannah River showed that river water is not being used for purposes of drinking or irrigation for at least 100 miles downriver (discussed in Section 4.1).

4-20

Figure 4.6-1 Average Annual H-3 Concentration in River Water 3500 3000 5

200 C 2000 0

• 8 1ooo00 0

Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 Year Indicator -U--Control Special -

MDC 4-21 cCc

Table 4.6-1 Average Annual H-3 Concentration in River Water Year Special Indicator Control Difference MDD Annual Site (pCi/1)

(pCiIl)

(pCi/1)

Between (pCi/1)

Tritium Indicator and Released Control (Ci)

(pc /o Pre-op 1900 650 665

-15 145 NA 1987 1411 680 524 156 416 321 1988 1430 843 427 416 271 390 1989 1268 1293 538 755 518 918 1990 1081 1142 392 750 766 1172 1991 1298 1299 828 471 626 1094 1992 929 1064 371 693 714 1481 1993 616 712 238 474 1526 761 1994 774 1258 257 1001 2009 1052 1995 699 597 236 361 766 968 1996 719 1187 387 800 2147 1637 1997 686 1547 254 1293 1566 1449 1998 640 1226 196 1030 1313 1669 1999 859 2005 389 1616 1079 1674 2000 885 1564 496 1068 1786 869 2001 931 2101 743 1358 1696 1492 4-22

4.7 Drinking Water Samples are collected at a control location (Station 80 - the Augusta Water Treatment Plant in Augusta, Georgia located about 56 river miles upriver), and at two indicator locations (Station 87 - the Beaufort-Jasper County Water Treatment Plant near Beaufort, South Carolina, 112 river miles downriver; and Station 88 the Cherokee Hill Water Treatment Plant near Port Wentworth, Georgia, 122 river miles downriver).

These upriver and downriver distances in river miles are the distances from the plant to the point on the river where water is diverted to the intake for each of these water treatment plants.

Water samples are taken near the intake of each water treatment plant (raw drinking water) using automatic samplers that take periodical small aliquots from the stream. These composite samples are collected monthly along with a grab sample of the processed water coming from the treatment plants (finished drinking water). Quarterly composites are made from these monthly collections for both raw and processed river water.

Gross beta and gamma isotopic analyses are performed on each of the monthly samples while tritium analysis is conducted on the quarterly composites. An 1-131 analysis is not required to be conducted on these samples, since the dose calculated from the consumption of water is less than 1 mrem per year (see ODCM Table 4-1). However, an 1-131 analysis is conducted on each of the monthly finished water grab samples, since a drinking water pathway exists.

Provided in Figures 4.7-1 and 4.7-2 and Tables 4.7-1 and 4.7-2, are the historical trends of the average gross beta concentrations found in the monthly collections of raw and finished drinking water.

For 2001, the indicator station average gross beta concentration in the raw drinking water was 0.27 pCi/l greater than the average gross beta concentration at the control station. This difference is not statistically discernible, since it is less than the MDD of 1.5 pCi/l.

For 2001, the indicator station average gross beta concentration in the finished drinking water was 2.67 pCi/l, which was 0.67 pCi/l greater than the average gross beta concentration at the control station. This difference is less than the MDD of 1.05 pCi/l and not statistically discernible. The gross beta concentrations at the indicator stations ranged from 0.45 to 6.2 pCi/l while the concentrations at the control station ranged from 1.06 to 4.1 pCi/l. Concentrations for the past four years are higher than gross beta results for finished drinking water during previous years of plant operation. However, these concentrations are only slightly higher than gross beta concentrations found during preoperation. Further, the concentration of 2.67 pCi/l is less than the required MDC of 4.0 pCi/l. There is no RL for gross beta in drinking water.

4-23

Figure 4.7-1 Table 4.7-1 Average Monthly Gross Beta Concentration in Water Raw Drinking Period Indicator Control

/1(pC)

(pCI Pre-op 2.70 1.90 1987 2.20 5.50 1988 2.67 3.04 1989 2.93 3.05 1990 2.53 2.55 1991 2.83 3.08 1992 2.73 2.70 1993 3.17 2.83 1994 3.51 3.47 1995 3.06 4.90 1996 5.83 3.02 1997 2.93 2.94 1998 3.31 2.58 1999 4.10 4.37 2000 4.52 3.59 2001 3.21 2.94 4-24 CIO

Figure 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water 4.5 3.5 3

2.5

1.

5-0.5 0E11 00 01 98 99 Po 87 38 89 90 91 92 93 94 Year I Indicator

-U Control

-MDC Table 4.7-2 95 96 97 Average Monthly Gross Beta Concentration in Finished Drinking Water Period Indicator Control (pCi/l)

(pCil)

Pre-op 2.90 1.80 1987 2.10 1.80 1988 2.28 2.35 1989 2.36 2.38 1990 2.08 1.92 1991 1.90 1.53 1992 2.09 1.67 1993 2.23 2.30 1994 2.40 2.68 1995 2.74 2.32 1996 2.19 2.21 1997 2.38 1.77 1998 3.23 1.67 1999 3.23 3.21 2000 3.39 2.68 2001 2.67 2.00 4 25 CuI L,

0 L)

Figure 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water 4.5 3.5 3

2.5

1.

5-0.5 0E11 00 01 98 99 Po 87 38 89 90 91 92 93 94 Year I Indicator

-U Control

-MDC Table 4.7-2 95 96 97 Average Monthly Gross Beta Concentration in Finished Drinking Water Period Indicator Control (pCi/l)

(pCil)

Pre-op 2.90 1.80 1987 2.10 1.80 1988 2.28 2.35 1989 2.36 2.38 1990 2.08 1.92 1991 1.90 1.53 1992 2.09 1.67 1993 2.23 2.30 1994 2.40 2.68 1995 2.74 2.32 1996 2.19 2.21 1997 2.38 1.77 1998 3.23 1.67 1999 3.23 3.21 2000 3.39 2.68 2001 2.67 2.00 4 25 CuI L,

0 L)

As provided in Table 3-1, there were no positive results during 2001 for the radionuclides of interest from the gamma isotopic analysis of the monthly collections for both raw and finished drinking water. Only one positive result has been found since operation began. Be-7 was found at a concentration of 68.2 pCi/I in the sample collected for September 1987 at Station 87. During preoperation Be 7 was found in about 5% of the samples at concentrations ranging from 50 to 80 pCi/l.

The MDC assigned for Be-7 in water is 124 pCi/l.

Also during preoperation, Cs-134 and Cs-137 were detected in about 7% of the samples at concentrations on the order of their MDCs which are 15 and 18 pCi/l, respectively.

1-131 was detected in finished drinking water in 1997 at levels near the MDC.

This was the first occurrence for detecting 1-131 in finished drinking water since operation began. During preoperation, it was detected in only one of 73 samples at a concentration of 0.77 pCi/l at Port Wentworth. The MDC and RL for 1-131 in drinking water are land 2 pCi/I, respectively.

Figures 4.7-3 and 4.7-4 and Tables 4.7-3 and 4.7-4 provide historical trending for the average tritium concentrations found in the quarterly composites of raw and finished drinking water collected at the indicator and control stations. The tables also list the calculated differences between the indicator and control stations, and list the MDDs between these two station groups.

The graphs and tables show that the tritium concentrations in the drinking water samples, both raw and finished, have been gradually trending downward since 1988. The small increase in average concentrations at the indicator stations for 1991 and 1992 reflect the impact of the inadvertent release from SRS of 7,500 Ci of tritium to the Savannah River about 10 miles downriver of VEGP, in December 1991 (SRS release data was obtained from "Release of 7,500 Curies of Tritium to the Savannah River from the Savannah River Site", Georgia Department of National Resources, Environmental Protection Division, Environmental Radiation Program, January 1992).

The 2001 raw drinking water indicator station tritium was 889 pCi/I, which was 364 pCi/l greater than the concentration determined at the control station. Because there was only one positive sample at the control station, no MDD could be calculated; however, application of the modified Student's t test showed that there was no discernible difference between the indicator and control station concentrations. The 2001 raw drinking water indicator station tritium was less than approximately 37% of that found in preoperation samples and samples collected during the first three years of operation.

The finished drinking water tritium concentration at the indicator stations during 2001 was 1037 pCi/l, which was 521 pCi/l greater than that found at the control station. Because there was only one positive sample at the control station, no MDD could be calculated; however, application of the modified Student's t test showed that there was no discernible difference between the indicator and control station concentrations. The indicator station concentration of tritium in finished drinking water for 2001 was less than approximately 40% of the concentrations measured during preoperation and the first three years of operation.

4-26

Figure 4.7-3 Table 4.7-3 Avertnoe Anniial 1-1-3, Concentrationnin Raw Dlrinkinv Water Period Indicator Control Difference MDD (pci/i)

(pCi/I)

Between (pCi/l)

Indicator and Control (pCiII)

Pre-op 2300 400 1900 1987 2229 316 1913 793 1988 2630 240 2390 580 1989 2508 259 2249 1000 1990 1320 266 1054 572 1991 1626 165 1461 834 1992 1373 179 1194 353 1993 955 0

955 NA 1994 871 0

871 NA 1995 917 201 716 NA 1996 1014 207 807 151 1997 956 230 726 61 1998 791 160 631 NA 1999 908 0

908 NA 2000 1020 373 647 704 2001 889 525 1

364 NA 4-27 Average Annual H-3 Concentration in Raw Drinking Water 3000 2500 2000 1000 0

L) 500 PO.

A7 At 89 90 91 92 93 95 A9 97 98 99 00 01 Year I--*--Indicator ---

Control -MDC

Figure 4.7-4 Table 4.7-4 hn flnh,*n flrin rinn Wate=nr ev a

nnnuaA

• t-'

t etnnl n

st e

ltn aLISfl Ill JtZlniL'.-ti tllZllltI Period Indicator Control Difference MDD (pCi/l)

(pCi/l)

Between (pCi/l)

Indicator and Control (pCi/l)

Pre-op 2900 380 2520 1987 2406 305 2101 1007 1988 2900 270 2630 830 1989 2236 259 1977 627 1990 1299 404 895 1131 1991 1471 225 1246 647 1992 1195 211 984 427 1993 993 0

993 NA 1994 880 131 749 270 1995 847 279 568 NA 1996 884 168 716 NA 1997 887 221 666 383 1998 713 180 533 NA 1999 920 263 657 NA 2000 1043 251 792 833

2001, 1037 516 521 NA 4 28 C13 Average Annual H-3 Concentration in Finished Drinking Water 500 000 C2000 500 0

00 500 w---------------------------

Po 87 Be 89 90 91 92 93 94 95 96 97 98 99 00 01 Pc 87 88 8

90 1

92Year I -

Indicator

-U--Control

-MDC I-I

4.8 Fish Table 2-1 calls for the collection of at least one sample of any anadromous species of fish in the vicinity of the plant discharge during the spring spawning season, and for the semiannual collection of at least one sample of any commercially or recreationally important species in the vicinity of the plant discharge area and in an area not influenced by plant discharges. Table 2-1 specifies that a gamma isotopic analysis be performed on the edible portions of each sample collected.

As provided in Table 2-2, a 5-mile stretch of the river is generally needed to obtain adequate fish samples.

For the semiannual collections, the control location (Station 81) extends from approximately 2 to 7 miles upriver of the plant intake structure, and the indicator location (Station 85) extends from about 1.4 to 7 miles downriver of the plant discharge structure. For anadromous species, all collection points can be considered as indicator stations.

American shad was collected as the anadromous species on March 13, 2001. As in all but two previous years of operation, no radionuclides were detected in 2001 from the gamma isotopic analysis of the anadromous species during the spring spawning season. In 1987, as well as in 1991, Cs-137 was found in a single sample of American shad at concentrations of 10 and 12 pCi/kg-wet, respectively.

The dates and compositions of the semiannual catches at the indicator and control stations during 2001 are shown below.

Date Indicator Control April 10 Channel Catfish Largemouth Bass Largemouth Bass Redear Sunfish October 09 Largemouth Bass White Catfish Largemouth Bass As indicated in Table 3-1, Cs-137 was the only radionuclide found in the semiannual collections of a commercially or recreationally important species of fish. It has been found in all but 3 of the 108 samples collected during operation and in all but 5 of the 32 samples collected during preoperation. As provided in Table 3-1, the average concentration at the indicator station was 8.5 pCi/kg-wet greater than that at the control station.

This difference is not statistically discernible, as it is less than the calculated MDD of 112 pCi/kg-wet. A discernible difference has not occurred for any year of operation or during preoperation.

Figure 4.8-1 and Table 4.8-1 provide the historical trending of the average concentrations of Cs-137 in units of pCi/kg-wet found in fish samples at the indicator and control stations. No trend is recognized in this data. The MDC and RL for Cs-137 in fish are 150 and 2000 pCi/kg-wet, respectively.

4-29

Figure 4.8-1 Table 4.8-1 Average Annual Cs-137 Concentration in Fish Year Indicator j

Control (pCi/kg-wet)

I (pCi/kg-wet)

Pre-op 590 340 1987 337 119 1988 66 116 1989 117 125 1990 103 249 1991 105 211 1992 178 80 1993 360 84 1994 165 200 1995 125 96 1996 194 404 1997 93 139 1998 190 200 1999 848 221 2000 55 96 2001 48 39 4 30

The only other radionuclide found in fish samples during operation is 1-131.

In 1989, it was found in one sample at the indicator station at a concentration of 18 pCi/kg-wet. In 1990, it was found in one sample at the indicator station and in one sample at the control station, at concentrations of 13 and 12 pCi/kg-wet, respectively. The MDC assigned to 1-131 in fish is 53 pCi/kg-wet.

During preoperation, Cs-134 was found in two of the 17 samples collected at the control station at concentrations of 23 and 190 pCi/kg-wet. The MDC and RL for Cs-134 are 130 and 1000 pCi/kg-wet, respectively. Nb-95 was also found in one of the control station samples at a concentration of 34 pCi/kg-wet. The assigned MDC and calculated RL for Nb-95 are 50 and 70,000 pCi/kg-wet, respectively.

4-31

4.9 Sediment Sediment was collected along the shoreline of the Savannah River on April 10 and October 2, 2001 at Stations 81 and 83. Station 81 is a control station located about 2.5 miles upriver of the plant intake structure while Station 83 is an indicator station located about 0.6 miles downriver of the plant discharge structure.

A gamma isotopic analysis was performed on each sample. The two radionuclides identified in 2001 samples were Be-7and Cs-137.

For Be-7, the average level at the indicator station during 2001 was 917 pCi/kg-dry less than that at the control station. Because there was only one positive sample at the control station, no MDD could be calculated; however a modified Student's t test showed that there is no discernible difference between the concentrations measured at the indicator station and the single value determined at the control station. There continues to be no statistically discernible difference between the indicator and control stations for Be-7 and on this basis its presence at the indicator station is not attributed to plant releases.

For Cs-137, the average concentration at the indicator station during 2001was 134.4 pCi/kg-dry greater than that at the control station. Because there was only one positive sample at the indicator station, no MDD could be calculated; however, application of Student's t test showed that there was no discernible difference between the indicator and the control stations. The Cs-137 level at the indicator station has averaged nearly 150 pCi/kg-dry greater than that at the control station over the entire period of operation. During preoperation, the Cs-137 was 170 pCi/kg-dry greater at the indicator station than at the control station.

The historical average concentrations of Be-7, Co-58, Co-60, and Cs-137 in sediment are plotted in Figures 4.9-1 through 4.9-4 along with listings of their concentrations in Tables 4.9-1 through 4.9-4. The concentrations of the solely man-made nuclides (Co-58, Co-60, & Cs-137) are consistent with past average concentrations. No pattern has been detected. Be-7, produced by man and nature, is also within the range that is typically seen.

During preoperation, Zr-95, Nb-95, Cs-134, and Ce-141 were detected in at least one of the control station samples and Nb-95 was detected in one of the indicator station samples. Be-7 and Cs-137 were found in several of the samples. The concentrations of these preoperational nuclides were on the order of their respective MDC values. The presence of these preoperational nuclides could be attributed to atmospheric weapons testing and the Chernobyl incident.

Mn-54 and 1-131 were found sporadically over several years of operation.

A summary of the positive results for these nuclides along with their applicable MDCs is provided in Table 4.9-5.

4-32

Figure 4.9-1 Table 4.9-1 Average Annual Be-7 Concentration in Sediment MDC=655 pCi/kg-dry Year Indicator Control

_(pCi/kg-dr~y)

(pCi/kg-dry)

Pre-op 1987 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 580 987 970 1300 465 826 2038 711 1203 1865 500U 543 810 415 545 42/

fý380 902 964 1155 1925 831 1130 1028 1396 662 1526 1697 1016 769 33214 2014 4-33 Average Annual Be-7 Concentration in Sediment 3500 3000 V\\

500 1500 0

Po 87 Be 89 90 91 92 93 94 95 96 97 98 99 00 01 Year S--findicator Control MDC

Figure 4.9-2 Table 4.9-2 Average Annual Co-58 Concentration in Sediment Year Indicator Control (pC kg-dry)

(pCi/kg-dry)

Pre-op 0

0 1987 0

0 1988 190 0

1989 135 0

1990 140 0

1991 0

0 1992 124 0

1993 0

0 1994 18.4 0

1995 42.4 0

1996 274 0

1997 0

0 1998 0

0 1999 0

0 2000 0

0 2001 0

0 4-34 I1(9 MDC-43 pCi/kg-dry

Figure 4.9-3 Table 4.9-3 Average Annual Co-60 Concentration in Sediment MDC=70 pCilkg-dry Year Indicator Control (pCi/kg-dry)

(pCilkg-dry)

Pre-op 0

0 1987 0

0 1988 62 0

1989 46 0

1990 46 0

1991 113 0

1992 59.5 0

1993 65.9 0

1994 85.2 0

1995 267 0

1996 344 0

1997 86 0

1998 263 0

1999 49.5 0

2000 131.3 0

2001 0

0 4-35 Cl-7 Average Annual Co-GO Concentration in Sediment 400 350 v300 250

0.

C20-----------------------------------

)

P.8 e8 09 2YC~r9 5

69 89 00

-4Indicator Control MDC]

Figure 4.9-4 Table 4.9-4 Average Annual Cs-137 Concentration in Sediment Year Indicator Control S(pCi/kg)

(pCi/kg)

"Pre-op 320 150 1987 209 111 1988 175 175 1989 230 125 1990 155 140 1991 246 100 1992 259 111 1993 345 115 1994 240 118 1995 357 123 1996 541 93 1997 184 98 1998 316 122 1999 197 97 2000 138 218 2001 252 118 4 36 Average Annual Cs-137 Concentration in Sediment 600 00 C300 20 0

10 F

(.)~/ 0/

Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 Year

-Indicator Control MDC MDC=180 pCi/kg

Table 4.9-5 Additional Sediment Nuclide Concentrations 4-37 Nuclide YEAR Indicator Control MDC

_______I j

(pCLkg-dry)

(pCilkg-dry)

(pCi/kg-dry)

Mn-54 1988 22 0

1989 18 0

42 1994 32 0

1-131 1992 194 20 53 1994 51 41

5.0 INTERLABORATORY COMPARISON PROGRAM In accordance with ODCM 4.1.3, the EL participates in an ICP which satisfies the requirements of Regulatory Guide 4.15, Revision 1, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment", February 1979.

The guide indicates that the ICP is to be conducted with the EPA (Environmental Protection Agency) Environmental Radioactivity Laboratory Intercomparison Studies (Cross-check) Program or an equivalent program, and that the ICP should include all of the determinations (sample medium/radionuclide combinations) that are offered by the EPA and included in the REMP.

The ICP is conducted by Analytics, Inc. of Atlanta, Georgia. Analytics has a documented QA (Quality Assurance) program and the capability to prepare QC (Quality Control) materials traceable to the National Institute of Standards and Technology. The ICP is a third party blind testing program which provides a means to ensure independent checks are performed on the accuracy and precision of the measurements of radioactive materials in environmental sample matrices.

Analytics supplies the crosscheck samples to the EL which performs the laboratory analyses in a normal manner.

Each of the specified analyses is performed three times. The results are then sent to Analytics who performs an evaluation which may be helpful to the EL in the identification of instrument or procedural problems.

The samples offered by Analytics and included in the EL analyses are gross beta in air filters, gamma-emitting radionuclides in air filters, gamma-emitting radionuclides in milk, gross beta in water, tritium in water, and gamma-emitting radionuclides in water. Normally, all of these types of samples are supplied each year by Analytics and analyzed by EL. In 2001, milk was offered by Analytics but, due to administrative problems, was not supplied to EL for analysis. To prevent recurrence of this situation, EL has established a long-term agreement with Analytics that includes provisions for supplying to EL all of the sample types listed above.

The accuracy of each result is measured by the normalized deviation, which is the ratio of the reported average less the known value to the total error. The total error is the square root of the sum of the squares of the uncertainties of the known value and of the reported average. The uncertainty of the known value includes all analytical uncertainties (counting statistics, calibration uncertainties, chemical yield etc.). The uncertainty of the reported average is the standard deviation of the analysis results performed by the EL. The precision of each result is measured by the coefficient of variation, which is defined as the standard deviation divided by the reported average. An investigation is undertaken whenever the absolute value of the normalized deviation is greater than three or whenever the coefficient of variation is greater than 15% for all radionuclides other than Cr-51 and Fe-59. For Cr-51 and Fe-59, an investigation is undertaken when the coefficient of variation exceeds the values shown as follows:

5-1

Nuclide Concentration

• Total Sample Activity Percent Coefficient (pCi) of Variation Cr-51

<300 NA 25 Cr-51 NA

>1000 25 Cr-51

>300

<1000 15 Fe-59

<80 NA 25 Fe-59

>80 NA 15

• For air filters, concentration units are pCi/filter.

For all other media, concentration units are pCi/liter.

As required by ODCM 4.1.3.3 and 7.1.2.3, a summary of the results of the EL's participation in the ICP is provided in Table 5-1 for: the gross beta analysis of an air filter; the gamma isotopic analysis of an air filter and water samples; and the tritium analysis of water samples.

Delineated in this table for each of the media/analysis combinations, are:

the specific radionuclides; Analytics' preparation dates; the known values with their uncertainties supplied by Analytics; the reported averages with their standard deviations; and the resultant normalized deviations and coefficients of variation expressed as a percentage.

It may be seen from Table 5-1 that all results were acceptable for precision, with one exception. The analysis of 1-131 in a water sample prepared on 06/14/2001, exceeded the coefficient of variation acceptance criterion of 15%. None of the analysis results exceeded the acceptance criteria for accuracy, which is a normalized deviation no greater than three. The outcome of the investigation into the result that failed to meet ICP acceptance criteria is provided in the following paragraph.

The precision deviation was from the determination of 1-131 in water by gamma spectroscopy. The precision result was outside the upper control limit. The high error of the gamma spectroscopy values was due to the low level of activity, approximately 14 pCi of 1-131, contained in the sample on the count date.

Although this level of activity is measurable by gamma spectroscopy as shown in the accuracy results, the low activity level presents high counting errors. This result was not due to sample processing nor analysis problems, therefore no further action will be necessary to address analytical problems. The cause of the low activity that resulted in the precision deviation was administrative in nature.

The water sample containing 1-131 was prepared by Analytics, but was not shipped to EL in a timely manner due to administrative problems. To reduce potential of recurrence of this problem, EL has established a long-term agreement with Analytics including provisions to assure that samples containing 1-131 are supplied to EL in a timely manner following preparation, including notification to EL prior to shipment for processing.

5-2

TABLE 5-1 (SHEET 1 of 2)

INTERLABORATORY COMPARISON PROGRAM RESULTS GROSS BETA ANALYSIS OF AN AIR FILTER (pCi/filter)

Analysis or Date Reported Known Standard Uncertainty of Percent Coef Normalized Radionuclide Prepared Average Value Deviation Known (3S) of Variation Deviation Gross Beta 12/06/01 114 106 3.71 1.67 3.25 1.97 GAMMA ISOTOPIC ANALYSIS OF AN AIR FILTER (pCi/filter)

Analysis or Date Reported Known Standard Uncertainty of Percent Coef Normalized Radionuclide Prepared Average Value Deviation Known (3S) of Variation Deviation Ce-141 12/06/01 288 314 7.23 5.33 2.51

-2.89 Co-58 12/06/01 273 293 4.77 5

1.75

-2.89 Co-60 12/06/01 188 171 7.91 3

4.21 2.01 Cr-51 12/06/01 402 412 30.31 7

7.54

-0.32 Cs-134 12/06/01 166 165 4.61 2.67 2.77 0.19 Cs-137 12/06/01 266 263 8.70 4.33 3.27 0.31 Fe-59 12/06/01 74 75 7.45 1.33 10.07

-0.13 Mn-54 12/06/01 125 123 6.17 2

4.94 0.31 Zn-65 12/06/01 95 84 12.03 1.33 12.66 0.91 GROSS BETA ANALYSIS OF WATER SAMPLE (pCi/liter)

Analysis or Date Reported Known Standard Uncertainty of Percent Coef Normalized Radionuclide Prepared Average Value Deviation Known (3S) of Variation Deviation Gross Beta 03/22/01 313 268 4.43 4.33 10.00 1.44 06/14/01 263 248 3.98 4.00 2.00 2.49

TABLE 5-1 (SHEET 2 of 2)

INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES (pCi/liter)

Analysis or Date Reported Known Standard Uncertainty of Percent Coef Normalized Radionuclide Prepared Average Value Deviation Known (3S) of Variation Deviation Ce-141 03/22/01 91.0 94 8.22 1.67 9.04

-0.36 06/14/01 239.0 234 15.36 4.00 6.43 0.31 Co-58 03/22/01 51.5 48 5.40 0.67 10.39 0.73 06/14/01 113.0 139 8.70 2.33 7.70

-2.89 Co-60 03/22/01 152.0 147 6.42 2.33 4.22 0.73 06/14/01 201.0 194 7.65 3.33 3.80 0.84 Cr-51 03/22/01 242.0 242 41.31 4.00 17.07 0.00 06/14/01 327.0 322 64.26 5.33 19.65 0.08 Cs-134 03/22/01 115.0 129 13.25 2.00 11.52

-1.04 06/14/01 169.0 193 13.74 3.33 8.13

-1.70 Cs-137 03/22/01 99.0 102 7.33 1.67 7.40

-0.40 06/14/01 183.0 174 9.29 3.00 5.08 0.92 Fe-59 03/22/01 95.2 84 9.82 1.33 10.34 1.11 06/14/01 136.0 126 12.67 2.00 9.32 0.78 1-131 03/22/01 93.3 90 9.39 1.67 10.10 0.31 06/14/01 91.0 74 28.40 1.33 31.21 0.60 Mn-54 03/22/01 106.0 101 7.65 1.67 7.21 0.64 06/14/01 218.0 216 9.76 3.67 4.48 0.19 Zn-65 03/22/01 195.0 186 8.11 3.00 8.11 0.56 06/14/01 291.0 261 18.70 4.33 6.43 1.56 TRITIUM ANALYSIS OF WATER SAMPLES (pCi/liter)

6.0 CONCLUSION

S This report confirms the licensee's conformance with the requirements of Chapter 4 of the ODCM during 2001. It provides summaries of data collection activities and a discussion of the results of the laboratory analyses for the samples.

No discernible radiological impact upon the environment or the public as a consequence of plant discharges to the atmosphere and to the river was established for any REMP samples collected for 2001.

6-1

OFFSITE DOSE CALCULATION MANUAL FOR SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT Revision 19 November 28, 2001

VEGP ODCM DISTRIBUTION LIST Copy Number 1

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10 12 14 15 16 18 19 20 21 22 23 24 28 29 Manual Holder George R. Frederick Training Center Simulator VEGP Main Control Room Podium Nancy D. Lacey Manager, Engineering Support Primary Chemistry Lab Manager, Health Physics & Chemistry Manager, Training & Emergency Preparedness James A. Bailey W. R. Nicholson VEGP Document Control SAER Supervisor Chemistry Superintendent Joe B. Sills Dwight Hostetter Michael C. Nichols Shan Sundaram Terry G. Lamb Steve White s:\\vogtle\\odcm\\dI.doc Rev. 19 i

VEGP ODCM REVISION LOG Revision Number 0

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9 10 11 12 13 14 15 16 17 18 19 s:\\vogtle\\odcmVeVision Iog.doc ii Rev. 19 Date 8/85 8/86 11/86 3/87 7/87 10/87 9/88 3/91 2/93 1/94 2/96 12/96 2/97 6/97 12/97 1/99 7/99 10/00 8/01 11/01 Rev. 19 ii

VEGP ODCM TABLE OF CONTENTS PAGE D IST R IBU T IO N LIST................................................................................................................

R EV IS IO N LO G..........................................................................................................................

ii TABLE O F C O NTENTS.............................................................................................................

iii LIST O F TA B LES....................................................................................................................

viii LIST O F FIG URES............................................................................................................

x R E FE R EN C E S..........................................................................................................................

xi CHAPTER 1: INTRO DUCTION.............................................................................................

1-1 CHAPTER 2: LIQUID EFFLUENTS.......................................................................................

2-1 2.1 LIMITS OF OPERATION 2-1 2.1.1 Liquid Effluent Monitoring Instrumentation Control 2-1 2.1.2 Liquid Effluent Concentration Control 2-7 2.1.3 Liquid Effluent Dose Control 2-10 2.1.4 Liquid Radwaste Treatment System Control 2-11 2.1.5 Maior Chanqes to Liquid Radioactive Waste Treatment Systems 2-12 2.2 LIQUID RADWASTE TREATMENT SYSTEM 2-13 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2-17 2.3.1 General Provisions Regarding Setpoints 2-17 2.3.2 Setpoints for Radwaste System Discharge Monitors 2-19 2.3.3 Setpoints for Monitors on Normally Low-Radioactivity Streams 2-25 2.4 LIQUID EFFLUENT DOSE CALCULATIONS 2-26 2.4.1 Calculation of Dose 2-26 2.4.2 Calculation of Ai, 2-27 2.4.3 Calculation of CFN, 2-28 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2-37 2.5.1 Thirty-One Day Dose Proiections 2-37 2.5.2 Dose Proiections for Specific Releases 2-37 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS 2-38 CHAPTER 3: GASEOUS EFFLUENTS.................................................................................

3-1 3.1 LIMITS OF OPERATION 3-1 3.1.1 Gaseous Effluent Monitoring Instrumentation Control 3-1 3.1.2 Gaseous Effluent Dose Rate Control 3-7 iii Rev. 18

VEGP ODCM TABLE OF CONTENTS (continued)

PAGE 3.1.3 Gaseous Effluent Air Dose Control 3-10 3.1.4 Control on Gaseous Effluent Dose to a Member of the Public 3-11 3.1.5 Gaseous Radwaste Treatment System Control 3-12 3.1.6 Maior Changes to Gaseous Radioactive Waste Treatment Systems 3-13 3.2 GASEOUS WASTE PROCESSING SYSTEM 3-14 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3-20 3.3.1 General Provisions Reqardinq Noble Gas Monitor Setpoints 3-20 3.3.2 Setpoint for the Final Noble Gas Monitor on Each Release Pathway 3-22 3.3.3 Setpoints for Noble Gas Monitors on Effluent Source Streams 3-25 3.3.4 Determination of Allocation Factors, AG 3-27 3.3.5 Setpoints for Noble Gas Monitors with Special Requirements 3-29 3.3.6 Setpoints for Particulate and Iodine Monitors 3-29 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3-30 3.4.1 Dose Rates at and Beyond the Site Boundary 3-30 3.4.2 Noble Gas Air Dose at or Beyond Site Boundary 3-31 3.4.3 Dose to a Member of the Public at or Beyond Site Boundar 3-35 3.4.4 Dose Calculations to Support Other Requirements 3-38 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3-44 3.5.1 Thirty-One Day Dose Proiections 3-44 3.5.2 Dose Proiections for Specific Releases 3-45 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS 3-46 CHAPTER 4:

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM................... 4-1 4.1 LIMITS OF OPERATION 4-1 4.1.1 Radiological Environmental Monitoring 4-1 4.1.2 Land Use Census 4-9 4.1.3 Interlaboratory Comparison Program 4-10 4.2 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS 4-11 CHAPTER 5: TOTAL DOSE DETERMINATIONS.................................................................

5-1 5.1 LIMIT OF OPERATION 5-1 5.1.1 Aoglicability 5-1 5.1.2 Actions 5-1 5.1.3 Surveillance Reguirements 5-1 5.1.4 Basis 5-1 5.2 DEMONSTRATION OF COMPLIANCE 5-3 iv Rey. 18 Rev. 18 iv

VEGP ODCM TABLE OF CONTENTS (continued)

PAGE CHAPTER 6:

POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY..................................................

6-1 6.1 REQUIREMENT FOR CALCULATION 6-1 6.2 CALCULATIONAL METHOD 6-1 C HAPTER 7: REPO RTS.......................................................................................................

7-1 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT 7-1 7.1.1 Requirement for Report 7-1 7.1.2 Report Contents 7-1 7.2 RADIOACTIVE EFFLUENT RELEASE REPORT 7-3 7.2.1 Requirement for Report 7-3 7.2.2 Report Contents 7-3 7.3 MONTHLY OPERATING REPORT 7-6 7.4 SPECIAL REPORTS 7-6 CHAPTER 8: METEOROLOGICAL MODELS.......................................................................

8-1 8.1 ATMOSPHERIC DISPERSION 8-1 8.1.1 Ground-Level Releases 8-1 8.1.2 Elevated Releases 8-2 8.1.3 Mixed-Mode Releases 8-4 8.2 RELATIVE DEPOSITION 8-5 8.2.1 Ground-Level Releases 8-5 8.2.2 Elevated Releases 8-5 8.2.3 Mixed-Mode Releases 8-6 8.3 ELEVATED PLUME DOSE FACTORS 8-6 CHAPTER 9: METHODS AND PARAMETERS FOR CALCULATION OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, R8 1pj.........................

..................... 9-1 9.1 INHALATION PATHWAY FACTOR 9-1 9.2 GROUND PLANE PATHWAY FACTOR 9-2 9.3 GARDEN VEGETATION PATHWAY FACTOR 9-3 9.4 GRASS-COW-MILK PATHWAY FACTOR 9-6 9.5 GRASS-GOAT-MILK PATHWAY FACTOR 9-9 9.6 GRASS-COW-MEAT PATHWAY FACTOR 9-12 CHAPTER 10: DEFINITIONS OF EFFLUENT CONTROL TERMS.......................................

10-1 10.1 TERMS SPECIFIC TO THE ODCM 10-1 10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS 10-5 v

Rev. 18

VEGP ODCM LIST OF TABLES Table 2-1.

Table 2-2.

Table 2-3.

Table 2-4.

Table 2-5.

Table 2-6.

Table 2-7.

Table 2-8.

Table 3-1.

Table 3-2.

Table 3-3.

Table 3-4.

Table 3-5.

Table 3-6.

Table 3-7.

Table 3-8.

Table 3-9.

Table 3-10.

Table 3-11.

Table 3-12.

Table 4-1.

Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Liquid Waste Sampling and Analysis Program Applicability of Liquid Monitor Setpoint Methodologies Parameters for Calculation of Doses Due to Liquid Effluent Releases Element Transfer Factors Adult Ingestion Dose Factors Site-Related Ingestion Dose Factors, Aj, Radioactive Gaseous Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Gaseous Waste Sampling and Analysis Program Applicability of Gaseous Monitor Setpoint Methodologies Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated Finite Plume Attributes of the Controlling Receptor Raipj for Ground Plane Pathway, All Age Groups R8 Ipj for Inhalation Pathway, Child Age Group RaIpj for Inhalation Pathway, Adult Age Group R,,pj for Cow Meat Pathway, Child Age Group Raipj for Garden Vegetation, Child Age Group Radiological Environmental Monitoring Program vi Rev. 18 PAGE 2-3 2-5 2-8 2-18 2-30 2-31 2-32 2-35 3-3 3-5 3-8 3-21 3-33 3-34 3-37 3-39 3-40 3-41 3-42 3-43 4-3 Rev. 18 vi

Table 4-2.

Table 4-3.

Table 4-4.

Table 6-1.

Table 8-1.

Table 9-1.

Table 9-2.

Table 9-3.

Table 9-4.

Table 9-5.

Table 9-6.

Table 9-7.

Table 9-8.

Table 9-9.

Table 9-10.

Table 9-11.

Table 9-12.

Table 9-13.

Table 9-14.

Table 9-15.

VEGP ODCM LIST OF TABLES (Continued)

PAGE Reporting Levels for Radioactivity Concentrations in Environmental Samples Values for the Minimum Detectable Concentration (MDC)

Radiological Environmental Monitoring Locations Attributes of Member of the Public Receptor Locations Inside the Site Boundary Terrain Elevation Above Plant Site Grade Miscellaneous Parameters for the Garden Vegetation Pathway Miscellaneous Parameters for the Grass-Cow-Milk Pathway Miscellaneous Parameters for the Grass-Goat-Milk Pathway Miscellaneous Parameters for the Grass-Cow-Meat Pathway Individual Usage Factors Stable Element Transfer Data Inhalation Dose Factors for the Infant Age Group Inhalation Dose Factors for the Child Age Group Inhalation Dose Factors for the Teenager Age Group Inhalation Dose Factors for the Adult Age Group Ingestion Dose Factors for the Infant Age Group Ingestion Dose Factors for the Child Age Group Ingestion Dose Factors for the Teenager Age Group Ingestion Dose Factors for the Adult Age Group External Dose Factors for Standing on Contaminated Ground vii Rev. 18 4-7 4-8 4-12 6-2 8-7 9-5 9-8 9-11 9-14 9-15 9-16 9-17 9-20 9-23 9-26 9-29 9-32 9-35 9-38 9-41

VEGP ODCM LIST OF FIGURES Figure 2-1.

Figure 2-2.

Figure 2-3.

Figure 3-1.

Figure 3-2.

Figure 3-3.

Figure 3-4.

Figure 3-5.

Figure 3-6.

Figure 4-1.

Figure 4-2.

Figure 4-3.

Figure 4-4.

Figure 8-1.

Figure 8-2.

Figure 8-3.

Figure 8-4.

Figure 8-5.

Figure 8-6.

Figure 8-7.

Figure 8-8.

Figure 8-9.

Figure 8-10.

Unit 1 Liquid Radwaste Treatment System Unit 2 Liquid Radwaste Treatment System Liquid Radwaste Discharge Pathways Schematic Diagram of the Gaseous Radwaste Treatment System Schematic Diagram of the Unit 1 Plant Vent Release Pathway Schematic Diagram of the Unit 2 Plant Vent Release Pathway Schematic Diagram of the Turbine Building Vent Release Pathway (Typical of Both Units)

Schematic Diagram of the Dry Active Waste Processing Building Ventilation Release Pathway Schematic Diagram of the Radwaste Processing Facility Ventilation Release Pathway Terrestrial Stations Near Site Boundary Terrestrial Stations and Aquatic Stations, 0-5 Miles Terrestrial Stations Beyond 5 Miles Drinking Water Stations Vertical Standard Deviation of Material in a Plume (ar)

Terrain Recirculation Factor (Kr)

Plume Depletion Effect for Ground Level Releases Plume Depletion Effect for 30-Meter Releases Plume Depletion Effect for 60-Meter Releases Plume Depletion Effect for 100-Meter Releases Relative Deposition for Ground-Level Releases Relative Deposition for 30-Meter Releases Relative Deposition for 60-Meter Releases Relative Deposition for 100-Meter (or Greater) Releases viii Rev. 19 PAGE 2-14 2-15 2-16 3-15 3-16 3-17 3-18 3-19 3-19a 4-15 4-16 4-17 4-18 8-8 8-9 8-10 8-11 8-12 8-13 8-14 8-15 8-16 8-17

VEGP ODCM REFERENCES

1.

J.S. Boegli, R.R. Bellamy, W.L. Britz, and R.L. Waterfield, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," NUREG-01 33, October 1978.

2.

"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. NRC Regulatory Guide 1.109, March 1976.

3.

"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. NRC Regulatory Guide 1.109, Revision 1, October 1977.

4.

"Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Regulatory Guide 1.111, March 1976.

5.

"Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Regulatory Guide 1.111. Revision 1, July 1977.

6.

"Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," U.S. NRC Regulatory Guide 1.113, April 1977.

7.

W.R. Stokes III, T.W. Hale, J.L. Pearman, and G.R. Buell, "Water Resources Data, Georgia, Water Year 1983," U.S. Geological Survey Water Data Report GA-83-1, June 1984.

8.

Direct communication with the Water Resources Division, U.S. Geologoical Survey, U.S.

Department of the Interior, February 1985.

9.

Bernd Kahn, et aL, "Bioaccumulation of P-32 in Bluegill and Catfish," NUREG/CR-3981, February 1985.

10.

Memo from S.E. Ewald, Georgia Power Company, to C.C. Eckert, Georgia Power Company, May 9, 1988.

11.

Voatle Electric Generatinq Plant Units 1 and 2 Final Safety Analysis Report, Georgia Power Company.

12.

Voqtle Electric Generating Plant Units 1 and 2 Environmental Report - Operating License Stage, Georgia Power Company.

13.

Memo from A.C. Stalker, Georgia Power Company, to D.F. Hallman, Georgia Power Company, May 9, 1988.

14.

Vogtle Electric Generating Plant Land Use Survey - 1988, Georgia Power Company, April 1988.

ix Rev. 18 Rev. 18 ix

VEGP ODCM

15.

Letter to Southern Company Services from Pickard, Lowe, and Garrick, Inc., Washington, D.C., April 27, 1988.

16.

Letter to Bill Ollinger, Georgia Power Company, from T.L. Broadwell, Georgia Power Company, Atlanta, Georgia, June 22, 1988.

17.

Letter to Bill Ollinger, Georgia Power Company, from R.D. Just, Georgia Power Company, Atlanta, Georgia, July 8, 1988.

18.

L.A. Currie, Lower Limit of Detection: Definition and Elaboration of a Proposed Position of Radiological Effluent and Environmental Measurements, U.S. NRC Report NUREG/CR-4007, 1984.

19.

"Radiological Assessment Branch Technical Position," U.S. Nuclear Regulatory Commission, Revision 1, November 1979.

20.

D.C. Kocher, "Radioactive Decay Data Tables," U.S. DOE Report DOE/TIC-1 1026, 1981.

21.

J.E. Till and H.R. Meyer, eds., Radiological Assessment, U.S. NRC Report NUREG/CR 3332, 1983.

22.

Letter to Bill Ollinger, Southern Nuclear Operating Company, from Gary D. Johnson, Georgia Power Company, December 21, 1995.

x Rev. 18

VEGP ODCM CHAPTER 1 INTRODUCTION The Offsite Dose Calculation Manual is a supporting document of the Technical Specifications.

As such, it describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous effluents, and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm setpoints. In addition, it contains the following:

The controls required by the Technical Specifications, governing the radioactive effluent and radiological environmental monitoring programs.

Schematics of liquid and gaseous radwaste effluent treatment systems, which include designation of release points to UNRESTRICTED AREAS.

A list and maps indicating the specific sample locations for the Radiological Environmental Monitoring Program.

Specifications and descriptions of the information that must be included in the Annual Radiological Environmental Operating Report and the Radioactive Effluent Release Report required by the Technical Specifications.

The ODCM will be maintained at the plant for use as a reference guide and training document of accepted methodologies and calculations. Changes in the calculational methods or parameters will be incorporated into the ODCM in order to ensure that it represents current methodology in all applicable areas. Any computer software used to perform the calculations described will be maintained current with the ODCM.

Equations and methods used in the ODCM are based on those presented in NUREG-0133 (Reference 1), in Regulatory Guide 1.109 (References 2 and 3), in Regulatory Guide 1.111 (References 4 and 5), and in Regulatory Guide 1.113 (Reference 6).

1-1 Rev. 18 Rev. 18 1-1

VEGP ODCM CHAPTER 2 LIQUID EFFLUENTS 2.1 LIMITS OF OPERATION The following Liquid Effluent Controls implement requirements established by Technical Specifications Section 5.0. Terms printed in all capital letters are defined in Chapter 10.

2.1.1 Liquid Effluent Monitorinq Instrumentation Control In accordance with Technical Specification 5.5.4.a, the radioactive liquid effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits specified in Section 2.1.2 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with Section 2.3.

2.1.1.1 Applicability This limit applies at all times.

2.1.1.2 Actions With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint to a conservative value.

With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-1. Restore the INOPERABLE instrumentation to OPERABLE status within 30 days, or if unsuccessful, explain in the next Radioactive Effluent Release Report, per Technical Specification 5.6.3, why this inoperability was not corrected in a timely manner.

This control does not affect shutdown requirements or MODE changes.

2.1.1.3 Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 2-2. Specific instrument numbers are provided in parentheses for information only. The numbers apply to each unit. These numbers will help to identify associated channels or loops and are not intended to limit the requirements to the specific instruments associated with the number.

2-1 Rev. 18 Rev. 18 2-1

VEGP ODCM 2.1.1.4 Basis The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 2.3 to ensure that the alarm/trip will occur prior to exceeding the limits of Section 2.1.2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

Rev. 18 2-2

VEGP ODCM

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Rev. 18 2-3 I atle e"- i.

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OPERABILITY RequirementSa Instrument Minimum Channels Operable ACTION

1. Radwaste Monitors Providing Alarm and Automatic Termination of Release

a.

Liquid Radwaste Effluent Line (RE-0018) 1 37

b.

Steam Generator Blowdown Effluent Line 1

38 (RE-0021)

c.

Turbine Building Effluent Line (RE-0848) 1 38

2.

Radwaste Monitors Providing Alarm, but Not Automatic Termination of Release NSCW Effluent Line (RE-0020 A&B) 1 I

39

3.

Flowrate Measurement Devices

a.

Liquid Radwaste Effluent Line (FT-0018),

1 40 (FT-1035), or (FT-1036)

b.

Steam Generator Blowdown Effluent Line 1

40 (FT-0021)

c.

Flow to Blowdown Sump (AFQI-7620, F17620A) 1 40

a.

All requirements in this table apply to each unit.

VEGP ODCM Table 2-1 (contd).

Notation for Table 2 ACTION Statements ACTION 37 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a.

At least two independent samples are analyzed in accordance with Section 2.1.2.3, and

b.

At least two technically qualified members of the Facility Staff independently verify the discharge line valving and the release rate calculations.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 38 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross radioactivity at a MINIMUM DETECTABLE CONCENTRATION no higher than 1 x 10-7 p.CVmL using gross beta/gamma counting or 5 x 10-e Ci/mL for the principal gamma emitters using gamma-ray spectroscopy.

a.

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

b.

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

ACTION 39-With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for radioactivity at a MINIMUM DETECTABLE CONCENTRATION no higher than 1 x 10 liCVmL using gross beta/gamma counting or 5 x 10-e CVmL for the principal gamma emitters using gamma-ray spectroscopy.

ACTION 40 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves generated in place may be used to estimate flow.

2-4 Rev. 18 Rev. 18 2-4

VEGP ODCM Table 2-2.

Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Surveillance Requirementsd CHANNEL CHANNEL Instrument CHANNEL SOURCE CALIBRA-OPERATIONAL CHECK CHECK TION TEST

1. Radwaste Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RE-0018)

D P

Rb Qa(I)

b. Steam Generator Blowdown Effluent Line (RE-0021)

D M

Rb Qa(l)

c. Turbine Building Effluent Line (RE-0848)

D M

RbQa()

2.

Radwaste Monitors Providing Alarm, but Not Automatic Termination of Release NSCW Effluent Line R

(RE-0020 A&B)

D M

Rb Qa(2)

3.

Flowrate Measurement Devices

a. Liquid Radwaste Effluent Line (FT 0018), (FT 1084NB), or (FT 1085NB)

Dc NA R

NA

b. Steam Generator Blowdown Effluent Line (FT-0021)

Dc NA R

NA

c. Flow to Blowdown Sump (AFQI-7620, F17620A)

Dc NA R

Q Rev. 18 2-5

VEGP ODCM Table 2-2 (contd).

Notation for Table 2-2

a.

In addition to the basic functions of a CHANNEL OPERATIONAL TEST (Section 10.2):

(1)

The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room annunciation occurs (for item a. below only); and control room CRT indication occurs (if any of the following conditions exist):

(a)

Instrument indicates measured levels above the alarm/trip setpoint; (b)

Instrument indicates an "Equipment Trouble" alarm; (c)

Instrument indicates a "Low" alarm; or (d)

Instrument indicates channel "Deactivated".

(2)

The CHANNEL OPERATIONAL TEST shall also demonstrate that control room annunciation occurs (for item a. below only); and that control room CRT indication occurs (if any of the following conditions exist):

(a)

Instrument indicates measured levels above the alarm/trip setpoint; (b)

Instrument indicates an "Equipment Trouble" alarm; (c)

Instrument indicates a "Low" alarm; or (d)

Instrument indicates channel "Deactivated".

b.

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

c.

CHANNEL CHECK shall consist of verifying indication of flow during periods of release.

CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.

d.

All requirements in this table apply to each unit.

2-6 Rev. 18 Rev. 18 2-6

VEGP ODCM 2.1.2 Liquid Effluent Concentration Control In accordance with Technical Specifications 5.5.4.b and 5.5.4.c, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 1 x 1 0e gCi/mL total activity.

2.1.2.1 Applicability This limit applies at all times.

2.1.2.2 Actions With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the limits stated in Section 2.1.2, immediately restore the concentration to within the stated limits.

This control does not affect shutdown requirements or MODE changes.

2.1.2.3 Surveillance Requirements The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 2-3. The results of radioactive analyses shall be used with the calculational methods in Section 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2.

2.1.2.4 Basis This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe 135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2 (1959). The resulting concentration of 2 x 10 was then multiplied by the ratio of the effluent concentration limit for Xe-1 35, stated in Appendix B, Table 2, Column 1 of 10 CFR 20 (paragraphs 20.1001 to 20.2401), to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 (paragraphs 20.1 to 20.601), to obtain the limiting concentration of 1 x 10e p.CVmL.

2-7 Rev. 18 Rev. 18 2-7

VEGP ODCM Table 2-3.

Radioactive Liquid Waste Sampling and Analysis Program 2-8 Rev. 18 Sampling and Analysis Requirementsa'b MINIMUM DETECTABLE CONCENTRA Minimum TO MC Liquid Release Sampling Analysis Type of Activity TION (MDC)

Type FREQUENCY FREQUENCY Analysis (pgCi/mL)

A. BATCH RELEASES PRINCIPAL 5 E-7 GAMMA EMITTERS P

P Each BATCH Each BATCH 1-131 1 E-6

1. Waste Monitor Dissolved and 1 E-5 Tank P

Entrained Gases One BATCH/M M

(Gamma Emitters)

2. Drainage of H-3 1 E-5 System P

M Each BATCH COMPOSITE Gross Alpha 1 E-7 Sr-89, Sr-90 5 E-8 P

Q Each BATCH COMPOSITE Fe-55 1 E-6 B. CONTINUOUS RELEASES PRINCIPAL 5 E-7 GAMMA EMITTERS W

Continuous COMPOSITE 1-131 1 E-6 Dissolved and 1 E-5 M

Entrained Gases Grab Sample M

(Gamma Emitters)

Waste Water H-3 1 E-5 Retention Basinc M

Continuous COMPOSITE Gross Alpha 1 E-7 Sr-89, Sr-90 5 E-8 Q

Continuous COMPOSITE Fe-55 1 E-6 Rev. 18 2-8

VEGP ODCM Table 2-3 (contd).

Notation for Table 2-3

a.

All requirements in this table apply to each unit.

b.

Terms printed in all capital letters are defined in Chapter 10.

c.

The WWRB will not be considered a release point until there is a confirmed primary to secondary leak. Once a primary to secondary leak has been confirmed, this composite shall be analyzed as specified until the leak is repaired. This surveillance will continue until three consecutive weekly composite samples have shown no activity above the MDC.

Rev. 18 2-9

VEGP ODCM 2.1.3 Liquid Effluent Dose Control In accordance with Technical Specifications 5.5.4.d and 5.5.4.e, the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS shall be limited:

a.

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

b.

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

2.1.3.1 Applicability These limits apply at all times.

2.1.3.2 Actions With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the limits of Section 2.1.3., prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which identifies the cause(s) for exceeding the limit(s); defines the corrective actions to be taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the limits of Section 2.1.3.

This control does not affect shutdown requirements or MODE changes.

2.1.3.3 Surveillance Requirements At least once per 31 days, cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined, for each unit, in accordance with Section 2.4.

2.1.3.4 Basis This control is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The limits stated in Section 2.1.3 implement the guides set forth in Section II.A of Appendix I. The ACTIONS stated in Section 2.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculations in Section 2.4 implement the requirements in Section III.A of Appendix I, which state that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Section 2.4 for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the Rev. 18 2-10

VEGP ODCM methodology provided in Regulatory Guide 1.109 (Reference 3) and Regulatory Guide 1.113 (Reference 6).

This control applies to the release of liquid effluents from each unit at the site. The liquid effluents from shared LIQUID RADWASTE TREATMENT SYSTEMs are to be proportioned between the units.

2.1.4 Uquid Radwaste Treatment System Control In accordance with Technical Specification 5.5.4.f, the LIQUID RADWASTE TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the system shall be used to reduce radioactivity in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS would exceed 0.06 mrem to the total body or 0.2 mrem to any organ of a MEMBER OF THE PUBLIC in 31 days.

2.1.4.1 Applicability This limit applies at all times.

2.1.4.2 Actions With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the LIQUID RADWASTE TREATMENT SYSTEM not in operation, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which includes the following information:

a.

Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reason for the inoperability,

b.

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

c.

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

This control does not affect shutdown requirements or MODE changes.

2.1.4.3 Surveillance Requirements Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with Section 2.5, during periods in which the LIQUID RADWASTE TREATMENT SYSTEMs are not being fully utilized.

The LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE by meeting the controls of Sections 2.1.2 and 2.1.3.

2.1.4.4 Basis The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the UNRESTRICTED AREAS. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will Rev. 18 2-11

VEGP ODCM be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the LIQUID RADWASTE TREATMENT SYSTEM were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

This control applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

2.1.5 Maior Chanaes to Liquid Radioactive Waste Treatment Systems Licensee initiated MAJOR CHANGES TO LIQUID RADIOACTIVE WASTE TREATMENT SYSTEMS:

a.

Shall be reported to the Nuclear Regulatory Commission in the Radioactive Effluent Release Report for the period in which the change was implemented. The discussion of each change shall contain the information described in Section 7.2.2.7.

b.

Shall become effective upon review and approval by the General Manager -

Nuclear Plant.

Rev. 18 2-12

VEGP ODCM 2.2 LIQUID RADWASTE TREATMENT SYSTEM The Vogtle Electric Generating Plant is located on the west bank of the Savannah River approximately 151 river miles from the Atlantic Ocean. There are two pressurized water reactors on the site. Each unit is served by a separate LIQUID RADWASTE TREATMENT SYSTEM; however, certain components are shared between the two systems. Schematics of the LIQUID RADWASTE TREATMENT SYSTEMs are presented in Figure 2-1 and Figure 2-2. Liquid discharge pathways are shown in Figure 2-3.

All liquid radwastes treated by the LIQUID RADWASTE TREATMENT SYSTEM are collected in 5,000-gallon or 20,000-gallon waste monitor tanks. Releases from the waste monitor tanks are to the discharge line from the blowdown sump, and from there to the Savannah River. The blowdown sump also receives input from the waste water retention basins, turbine plant cooling water blowdown, and nuclear service cooling water blowdown. Additional dilution water is available from the cooling tower makeup water bypass line.

Although no significant quantities of radioactivity are expected in the nuclear service cooling water, the steam generator blowdown processing system, or the turbine building drain system, these effluent pathways are monitored as a precautionary measure. The monitors serving the latter two pathways provide for automatic termination of releases from these systems in the event that radioactivity is detected above predetermined levels. These two systems discharge to the waste water retention basin. Sampling and analysis of releases via all three of these pathways must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.

2-13 Rev. 18 Rev. 18 2-13

VEGP ODCM Figure 2-1.

Unit 1 Liquid Radwaste Treatment System 2-14 Rev. 19

VEGP ODCM from Laundry and Hot Shower Tank (Unit 1)

To Discharge Figure 2-2.

Unit 2 Liquid Radwaste Treatment System 2-15 Rev. 19 Rev. 19 2-15

VEGP ODCM Radioactivity Monitors N.

G 1(2) RE0020 G

1(2) RE0021 G

1(2) RE0848 G

1(2) RE0018

-Th bhowdown sump is common to both units.

Figure 2-3.

Liquid Radwaste Discharge Pathways 2-16 Rev. 18

VEGP ODCM 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2.3.1 General Provisions Regarding Setpoints Liquid monitor setpoints calculated in accordance with the methodology presented in this section will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower value for the high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint. If no release is planned for a particular pathway, or if there is no detectable activity in the planned release, the monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.

Two basic setpoint methodologies are presented below. For radwaste system discharge monitors, setpoints are determined to assure that the limits of Section 2.1.2 are not exceeded.

For monitors on streams that are not expected to contain significant radioactivity, the purpose of the monitor setpoints is to cause an alarm on low levels of radioactivity, and to terminate the release where this is possible. Section 2.1.1 establishes the requirements for liquid effluent monitoring instrumentation. Table 2-4 lists the monitors for which each of the setpoint methodologies is applicable.

2-17 Rev. 18 Rev. 18 2-17

VEGP ODCM Table 2-4.

Applicability of Liquid Monitor Setpoint Methodologies Liquid Radwaste Discharge Monitors Setpoint Method:

Section 2.3.2 Release Type:

BATCH Unit 1 or Unit 2 Liquid Waste Treatment System Effluent Monitor: 1 RE-0018/ 2RE-0018 Normally Low-Radioactivity Streams with Termination or Diversion upon Alarm Setpoint Method:

Section 2.3.3 Release Type:

CONTINUOUS Unit 1 or Unit 2 Steam Generator Blowdown Effluent Monitor: 1 RE-0021 / 2RE-0021 Unit 1 or Unit 2 Turbine Buildinq Drain Effluent Monitor: 1 RE-0848 / 2RE-0848 Normally Low-Radioactivity Streams with Alarm Only Setpoint Method:

Section 2.3.3 Release Type:

CONTINUOUS Unit 1 or Unit 2 Nuclear Service Cooling Water System Effluent Monitors (2 per unit): 1 RE-0020 A and B 2RE-0020 A and B 2-18 Rev. 18

VEGP ODCM 2.3.2 Setpoints for Radwaste System Discharge Monitors 2.3.2.1 Overview of Method LIQUID RADWASTE TREATMENT SYSTEM effluent line radioactivity monitors are intended to provide alarm and automatic termination of release prior to exceeding the limits specified in Section 2.1.2 at the point of release of the diluted effluent into the UNRESTRICTED AREA.

Therefore, their alarm/trip setpoints are established to ensure compliance with the following equation (equation adapted from Addendum to Reference 1):

"c. <TF.Cc (2.1)

F+f where:

CECL

=

the Effluent Concentration Limit corresponding to the mix of radionuclides in the effluent being considered for discharge, in giCVmL.

c

= the setpoint, in p.CVmL, of the radioactivity monitor measuring the concen tration of radioactivity in the effluent line prior to dilution and subsequent release. The setpoint represents a concentration which, if exceeded, could result in concentrations exceeding the limits of Section 2.1.2 in the UNRESTRICTED AREA.

f

=

the effluent flowrate at the location of the radioactivity monitor, in gpm.

F

=

the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA, in gpm. A predetermined dilution flowrate must be assured for use in the calculation of the radioactivity monitor setpoint.

TF

=

the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate effluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2; the tolerance factor must not exceed a value of 10.

While equation (2.1) shows the relationships of the critical parameters that determine the setpoint, it cannot be applied practically to a mixture of radionuclides with different Effluent Concentration Limits (ECLs). For a mixture of radionuclides, equation (2.1) is satisfied in a practicable manner based on the calculated ECL fraction of the radionuclide mixture and the dilution stream flowrate that can be assured for the duration of the release (Fd), by calculating the maximum permissible effluent flowrate (fmn) and the radioactivity monitor setpoint (c).

The setpoint method presented below is applicable to the release of only one tank of liquid radwaste per reactor unit at a given time. Liquid releases must be controlled administratively to ensure that this condition is met; otherwise, the setpoint method may not ensure that the limits of Section 2.1.2 are not exceeded.

2-19 Rev. 18 Rev. 18 2-19

VEGP ODCM 2.3.2.2 Setpoint Calculation Steps Steo 1:

Determine the radionuclide concentrations in the liquid waste being considered for release in accordance with the sampling and analysis requirements of Section 2.1.2.

All liquid radwastes treated by the LIQUID RADWASTE TREATMENT SYSTEM are collected in waste monitor tanks for sampling and analysis. The 5,000-gallon waste monitor tanks are recirculated for a minimum of 30 minutes, and the 20,000-gallon waste monitor tanks are recirculated for a minimum of 45 minutes. This mixing assures that a representative sample can be taken from the tank.

The total concentration of the liquid waste is determined by the results of all required analyses on the collected sample, as follows:

1c,=C+YC + Cf +Ct +

cg (2.2)

S g

where:

Ca

=

the gross concentration of alpha emitters in the liquid waste, not less than that measured in the most recent applicable composite sample.

Cs

=

the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste, not less than that measured in the most recent applicable composite sample.

Cf

=

the concentration of Fe-55 in the liquid waste, not less than that measured in the most recent applicable composite sample.

Ct

=

the concentration of H-3 in the liquid waste, not less than that measured in the most recent applicable composite sample.

Cg

=

the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the sample for the release under consideration.

The Cg term will be included in the analysis of each waste sample; terms for gross concentrations of alpha emitters, Sr-89, Sr-90, Fe-55, and tritium will be included in accordance with the sampling and analysis program required for the waste stream (see Section 2.1.2). For each analysis, only radionuclides identified and detected above background for the given measurement should be included in the calculation. When using the altemate setpoint methodology of step 5.b, the historical maximum values of Ca, Cý, Cf, and Ct shall be used.

Ste._2:

Determine the required dilution factor for the mix of radionuclides detected in the waste.

Measured radionuclide concentrations are used to calculate ECL fractions. The ECL fractions are used along with a safety factor to calculate the required dilution factor; this is the minimum ratio of dilution flowrate to waste flowrate that must be maintained throughout the release to ensure that the limits of Section 2.1.2 are not exceeded at the point of discharge into the UNRESTRICTED Rev. 18 2-20

VEGP ODCM AREA. The required dilution factor, RDF, is calculated as the sum of the dilution factors required for gamma emitters (RDF.T) and for non-gamma-emitters (RDF,):

PDF[ ECL,] [(sFXTF)]

(2.3)

=RDFY + RDF,*

,D ECL( ]

(2.4)

(SF XTF) where:

[. C'CS+

Cf

+ CL]

RDF,,Y ECLa.'7ECL, ELf ECU (2.5)

(SF XTF)

Ci

=

the measured concentration of radionuclide i as defined in step 1, in RCi/mL.

The Ca, Cs, Cf, and Ct terms will be included in the calculation as appropriate.

ECLi

=

the Effluent Concentration Umit for radionuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 (except for noble gases as discussed below).

In the absence of information regarding the solubility classification of a given radionuclide in the waste stream, the solubility class with the lowest ECL shall be assumed. For dissolved or entrained noble gases, the concentration shall be limited to lx10" pCVmL. For gross alpha, the ECL shall be 2x10.9 gCVmL; if specific alpha-emitting radionuclides are measured, the ECL for the specific radionuclide(s) should be used.

SF

=

the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1.

A value of 0.5 is reasonable for liquid releases; a more precise value may be developed if desired.

TF

=

the tolerance factor (as defined in Section 2.3.2.1).

Steo 3:

Determine the release-specific assured dilution stream flowrate.

Determine the dilution stream flowrate that can be assured during the release period, designated Fd; this value is the setpoint for the dilution stream flowrate measurement device.

If simultaneous radioactive releases are planned from the same or different reactor units, the dilution stream must be allocated among all the simultaneous releases. There will only be one such release per unit at a given time, unless there is detectable radioactivity in one of the normally low-radioactivity streams (see Section 2.3.3). Allocation of the dilution stream to multiple release paths is accomplished as follows:

Rev. 18 2-21

VEGP ODCM F,, = F(AFP)

(2.6) where:

Fd

=

the dilution flowrate allocated to release pathway p, in gpm.

AFp

=

the dilution allocation factor for release pathway p. AFp may be assigned any value between 0 and 1 for each active release pathway, under the condition that the sum of the AFp for all active release pathways for the entire plant site does not exceed 1.

Fd

=

the assured minimum dilution flow for the unit, in gpm.

In the normal case in which the only release pathways with detectable radioactivity are the LIQUID RADWASTE TREATMENT SYSTEMs of each unit, AFp for each unit may be assigned the value of 0.5 to permit releases from either unit to be made without regard to any releases from the other unit; if only one unit's LIQUID RADWASTE TREATMENT SYSTEM is releasing at a given time, its AFp may be increased proportionately. If more precise allocation factor values are desired, they may be determined based on the relative radiological impact of each active release pathway; this may be approximated by multiplying the RDF of each effluent stream by its respective planned release flowrate, and comparing these values. If only one simultaneous release is being made, its AFp may be assigned the value of 1, making F equal to Fd.

For the case where RDF < 1, the planned release meets the limits of Section 2.1.2 without dilution, and could be released with any desired effluent flowrate and dilution flowrate. However, in order to maintain individual doses due to liquid effluent releases as low as is reasonably achievable, no releases with detectable radioactivity should be made if the assured dilution flowrate, Fd, is less than 12,000 gpm.

Step 4:

Determine the maximum allowable waste discharge flowrate.

For the case where RDF > 1, the maximum permissible effluent discharge flowrate for this release pathway, fmp (in gpm), is calculated as follows:

Fd,,

(2.7)

(RDF-1)

For the case RDF _ 1, equation (2.7) is not valid. However, as discussed above, when RDF _ 1, the release may be made at full discharge pump capacity; the radioactivity monitor setpoint must still be calculated in accordance with Step 5 below.

NOTE 1:

Discharge flowrates are actually limited by the discharge pump capacity. When the calculated maximum permissible release flowrate exceeds the pump capacity, the release may be made at full capacity. Discharge flowrates less than the pump capacity must be achieved by throttling if this is available; if throttling is not available, the release may not be made as planned.

NOTE 2:

If, at the time of the planned release, there is detectable radioactivity due to plant operations in the dilution stream, the diluting capacity of the dilution stream is diminished. (In addition, sampling and analysis of the other radioactive effluents affecting the dilution stream must be sufficient to ensure that the liquid effluent 2-22 Rev. 18 Rev. 18 2-22

VEGP ODCM dose limits specified in the controls of Section 2.1.3 are not exceeded.) Under these conditions, equation (2.7) must be modified to account for the radioactivity present in the dilution stream prior to the introduction of the planned release:

fmp1 (RD-J"CELJ (2.8) where:

Cir

=

the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream.

fr

=

the effluent discharge flowrate of release pathway r.

If the entire dilution stream contains detectable activity due to plant operations, whether or not its source is identified, fr = Fd, and Cir is the concentration in the total dilution system. This note does not apply: a) if the RDF of the planned release is < 1; or b) if the release contributing radioactivity to the dilution stream has been accounted for by the assignment of an allocation factor.

Steop5:

Determine the maximum radioactivity monitor setpoint concentration.

Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 2.1.2 will not be exceeded.

Because the radioactivity monitor responds primarily to gamma radiation, the monitor setpoint cp for release pathway p (in gCVmL) is based on the concentration of gamma emitters in the waste stream, as follows:

cp = ApY Cg (2.9) g where:

Ap

=

an adjustment factor which will allow the setpoint to be established in a practical manner to prevent spurious alarms while allowing a margin between measured concentrations and the limits of Section 2.1.2.

Step 5.a.

If the concentration of gamma emitters in the effluent to be released is sufficient that the high alarm setpoint can be established at a level that will prevent spurious alarms, Ap should be calculated as follows:

1 AP =-

xADF

/RDF (2.10)

_ x (RFd

+ fp PDF f,.

2-23 Rev. 18 Rev. 18 2-23

VEGP ODCM where:

ADF

=

the assured dilution factor.

fa

=

the anticipated actual discharge flowrate for the planned release (in gpm), a value less than fmp. The release must then be controlled so that the actual effluent discharge flowrate does not exceed fap at any time.

Step 5.b.

Alternatively, Ap may be calculated as follows:

ADF - RDFr A,

=

RDFr (2.11)

Step 5.c.

Evaluate the computed value of Ap as follows:

If Ap _ 1, calculate the monitor setpoint, cp. However, if cp is within about 10 percent of Cg, it may be impractical to use this value of cp. This situation indicates that measured concentrations are approaching values which would cause limits of Section 2.1.2 to be exceeded.

Therefore, steps should be taken to reduce potential concentrations at the point of discharge; these steps may include decreasing the planned effluent discharge flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the effluent concentrations by further processing the liquid planned for release. Alternatively, allocation factors for the active liquid release pathways may be reassigned. When one or more of these actions has been taken, repeat Steps 1-5 to calculate a new radioactivity monitor setpoint.

If Ap < 1, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

2.3.2.3 Use of the Calculated Setpoint The setpoint calculated above is in the units CiCVmL. The monitor actually measures a count rate, subtracts a predetermined background count rate, and multiplies by a calibration factor to convert from count rate to lCi/mL.

Initial calibration of the monitors by the manufacturer and Georgia Power Company utilized NIST traceable liquid solutions with gamma ray emissions over the range 0.08 to 1.33 MeV, in the exact geometry of each production monitor. The calibration factor is a function of the radionuclide mix in the liquid to be released, and will be calculated for the monitor based on the results of the pre release sample results from the laboratory gamma-ray spectrometer system. The mix-dependent calibration factor will be used as the gain factor in the PERMS monitor, or used to modify the 2-24 Rev. 18 Rev. 18 2-24

VEGP ODCM calculated base monitor setpoint so that the default calibration factor in the PERMS monitor can be left unchanged.

Notwithstanding the initial calibration, monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to stream concentrations measured by liquid sample analysis. In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value.

2.3.3 Setpoints for Monitors on Normally Low-Radioactivity Streams Radioactivity in these streams (listed in Table 2-4 above) is expected to be at very low levels, generally below detection limits. Accordingly, the purpose of these monitors is to alarm upon the occurrence of significant radioactivity in these streams, and to terminate or divert the release where this is possible.

2.3.3.1 Normal Conditions When radioactivity in one of these streams is at its normal low level, its radioactivity monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.

2.3.3.2 Conditions Requiring an Elevated Setpoint Under the following conditions, radionuclide concentrations must be determined and an elevated radioactivity monitor setpoint determined for these pathways:

For streams that can be diverted or isolated, a new monitor setpoint must be established when it is desired to discharge the stream directly to the dilution water even though the radioactivity in the stream exceeds the level which would normally be diverted or isolated.

For streams that cannot be diverted or isolated, a new monitor setpoint must be established whenever: the radioactivity in the stream becomes detectable above the background levels of the applicable laboratory analyses; or the associated radioactivity monitor detects activity in the stream at levels above the established alarm setpoint.

When an elevated monitor setpoint is required for any of these effluent streams, it should be determined in the same manner as described in Section 2.3.2. However, special consideration must be given to Step 3. An allocation factor must be assigned to the normally low-radioactivity release pathway under consideration, and allocation factors for other release pathways discharging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the normally low-radioactivity streams must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.

2-25 Rev. 18 Rev. 18 2-25

VEGP ODCM 2.4 LIQUID EFFLUENT DOSE CALCULATIONS The following sub-sections present the methods required for liquid effluent dose calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D,, Aj, and CFN are summarized in Table 2-5.

2.4.1 Calculation of Dose The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a per-unit basis.

Therefore, the doses calculated in accordance with this section must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units.

For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid effluents released from each unit to UNRESTRICTED AREAS will be calculated as follows (equation from Reference 1, page 15):

D A=.[ X(AtICrIF]

(2.12) where:

DIC

=

the cumulative dose commitment to the total body or to any organ T, in mrem, due to radioactivity in liquid effluents released during the total of the m time periods At1.

Aft

=

the site-related adult ingestion dose commitment factor, for the total body or for any organ c, due to identified radionuclide i, in (mrem.mL)/(h.jCi).

Methods for the calculation of A, are presented below in Section 2.4.2. The values of Ak to be used in dose calculations for releases from the plant site are listed in Table 2-8.

At1

=

the length of time period 1, over which Cv and F, are averaged for liquid releases, in h.

CV

=

the average concentration of radionuclide i in undiluted liquid effluent during time period 1, in l.CVmL. Only radionuclides identified and detected above background in their respective samples should be included in the calculation.

F,

=

the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:

F (2.13) where:

ft=

the average undiluted liquid waste flowrate actually observed during the period of radioactivity release, in gpm.

2-26 Rev. 18

VEGP ODCM Ft

=

the average dilution stream flowrate actually observed during the period of radioactivity release, in gpm. If simultaneous releases from both units occur, the dilution stream flowrate Ft must be allocated between them. In such cases, F, is unit-specific.

Z

=

the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release, from Table 2-5.

NOTE:

In equation (2.13), the product (Ft x Z) is limited to 1000 cfs (=

448,000 gpm) or less. (Reference 1, Section 4.3.)

2.4.2 Calculation of A.

The site-related adult ingestion dose commitment factor, Ak, is calculated as follows (equation adapted from Reference 1, page 16, by addition of the irrigated garden vegetation pathway):

=

U.14x

, e Al

t. + UfBF e_',f + U CFv DFl Ar =(l.1'O VD-I'V 1.14x 1(

Uw Dw, tw Uf BFI tf Uv (2.14)

)5 = a units conversion factor, determined by:

106 pCVgCi x 103 m

+L

- 8760 h/y.

=

the adult drinking water consumption rate applicable to the plant site (L/y).

=

the dilution factor from the near field of the discharge structure for the plant site to the potable water intake location.

=

the decay constant for radionuclide i (he). Values of ý- used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 20.

=

the transit time from release to receptor for potable water consumption (h).

=

the adult rate of fish consumption applicable to the plant site (kg/y).

=

the bioaccumulation factor for radionuclide i applicable to freshwater fish in the receiving water body for the plant site, in (pCVkg)/(pCi/L) = (L/kg). For specific values applicable to the plant site, see Table 2-6.

=

the transit time from release to receptor for fish consumption (h).

=

the adult consumption rate for irrigated garden vegetation applicable to the plant site (kg/y).

Rev. 18 2-27 where:

VEGP ODCM CFý,

=

the concentration factor for radionuclide i in irrigated garden vegetation, as applicable to the vicinity of the plant site, in (pCi/kg)/(pCi/L). Methods for calculation of CF* are presented below in Section 2.4.3.

DF,

=

the dose conversion factor for radionuclide i for adults, in organ tr (mrem/pCi).

For specific values, see Table 2-7.

2.4.3 Calculation of Cf, The concentration factor for radionuclide i in irrigated garden vegetation, CFj, in (L/kg), is calculated as follows:

"* For radionuclides other than tritium (equation adapted from Reference 3, equations A-8 and A-9):

CF =M-Ir( J-e--,f.(-

1e-ý (2.15)

L J

"* For tritium (equation adapted from Reference 3, equations A-9 and A-10):

CFv = M. L, (2.16) where:

M

=

the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage.

I

=

the average irrigation rate during the growing season (L)/(m 2"h).

r

=

the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.

Yv

=

the areal density (agricultural productivity) of leafy garden vegetation (kg/m2) f,

=

the fraction of the year that garden vegetation is irrigated.

Bi,

=

the crop to soil concentration factor applicable to radionuclide i (pCi/kg garden vegetation)/(pCi/kg soil).

P

=

the effective surface density of soil (kg/m2).

X,

=

the decay constant for radionuclide i (h-'). Values of ^ý used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 20.

k.

=

the rate constant for removal of activity from plant leaves by weathering (h-1).

XE

=

the effective removal rate for activity deposited on crop leaves (h-1) calculated as: kEa = ?4 + Xw.

2-28 Rev. 18

VEGP ODCM te

=

the period of leafy garden vegetation exposure during the growing season (h).

tb

=

the period of long-term buildup of activity in soil (h).

th

=

the time between harvest of garden vegetation and human consumption (h).

the water content of leafy garden vegetation edible parts (L/kg).

2-29 Rev. 18

VEGP ODCM Table 2-5.

Parameters for Calculation of Doses Due to Liquid Effluent Releases Dose Calculation Receptor Locations:

Fish:

Drinking Water:

Vicinity of plant discharge 112 miles downstream, at Beaufort, SC (Reference 12)

Irrigated Garden Vegetation:

Numerical Parameters:

Parameter Value None (Reference 12)

Reference 10, for May through December 20, for January through April 730 L/y 8

48h 21 kg/y 24 h 0 kg/y 1.0+

No value 0.25 2.0 kg/m2 1.0+

240 kg/m2 0.0021 h1 (i.e., half-life of 14 d) 1440h (= 60 d) 1.31 x 105 h (= 15y) 24 h 0.92 11kg Ref. 11 Ref 3 Ref. 7 Ref. 3, Sec. A.2; Ref. 8 Ref. 3, Table E-5 Ref. 3, Sec. A.2 Ref. 12 Ref. 3, Table E-15.

Ref. 3, Table E-15 Ref. 3, Table E-15 Ref. 3, Table E-15 Ref. 3, Table E-15 Ref. 3, Table E-15 Ref. 3, Table E-15 Based on Ref. 21, Table 5.16 (for lettuce, cabbage, etc.)

-Because there is no irrigated garden vegetation pathway downstream of the plant site, the consumption of irrigated garden vegetation is set to zero, and the other pathway parameters are defaults.

+ -

There is no established default value for this parameter. The most conservative physically realistic value is 1.0.

2-30 z

Uw D,

tw Uf tf Uv M

I r

Yv fI P

te tb th 4*

Rev. 18

VEGP ODCM

Imn+

Trnfv

rtnr

i a U I e r -l l 1 IV "l L I I Q; lIlQ I

l I lu, Freshwater Fish Element BFi" H

9.0 E-01 C

4.6 E+03 Na 1.0 E+02 P

3.0 E+03 Cr 2.0 E+02 Mn 4.0 E+02 Fe 1.0 E+02 Co 5.0 E+01 Ni 1.0 E+02 Cu 5.0 E+01 Zn 2.0 E+03 Br 4.2 E+02 Rb 2.0 E+03 Sr 3.0 E+01 Y

2.5 E+01 Zr 3.3 E+00 Nb 5.5 E+02 Mo 1.0 E+01 Tc 1.5 E+01 Ru 1.0 E+01 Rh 1.0 E+01 Ag 2.3 E+00 Sb 2.0 E+02 Te 4.0 E+02 I

1.5 E+011 CS 2.0 E+03 Ba 4.0 E+00 La 2.5 E+011 Ce 1.0 E+00 Pr 2.5 E+01 Nd 2.5 E+011 W

1.2 E+03

-Np 1.0 E+011 Bioaccumulation Factors for freshwater fish, in (pCi/kg)/(pCi/L). They are obtained from Reference 3 (Table A-i), except as follows: Reference 9 for P; Reference 2 (Table A

8) for Ag; and Reference 10 for Nb and Sb.

Rev. 18 2-31

='if" i-,l rl

VEGP ODCM i aoie e-t.

PMUUIL IIIYWQLIUI V-;)-

Nuclide-f Bone LvrjT.

Body jhyroid Kidney~

cc.

1.

H-3 C-14 Na-24 P-32 Cr-51 Mn-54 Mn-56 Fe-55 Fe-59 Co-58 Co-60 Ni-63 Ni-65 Cu-64 Zn-65 Zn-69 Br-83 Br-84 Br-85 Rb-86 Rb-88 Rb-89 Sr-89 Sr-90 Sr-91 A

No Data 2.84E-06 1.70E-06 1.93E-04 No Data No Data No Data 2.75E-06 4.34E-06 No Data No Data 1.30E-04 5.28E-07 No Data 4.84E-06 1.03E-08 No Data No Data No Data No Data No Data No Data 3.08E-04 7.58E-03 1.05E-07 5.68E-07 1.70E-06 1.20E-05 No Data 4.57E-06 1.15E-07 1.90E-06 1.02E-05 7.45E-07 2.14E-06 9.01 E-06 6.86E-08 8.33E-08 1.54E-05 1.97E-08 No Data No Data No Data 2.11 E-05 6.05E-08 4.01 E-08 No Data No Data 5.67E-06 No Data 1.05E-07 5.68E-07 1.70E-06 7.46E-06 2.66E-09 8.72E-07 2.04E-08 4.43E-07 3.91 E-06 1.67E-06 4.72E-06 4.36E-06 3.13E-08 3.91 E-08 6.96E-06 1.37E-09 4.02E-08 5.21 E-08 2.14E-09 9.83E-06 3.21 E-08 2.82E-08 8.84E-06 1.86E-03 2.29E-07 No Data No Data No Data No Data All values are in (mrem/pCi ingested). They are obtained from Reference 3 (Table E-1 1),

except as follows: Reference 2 (Table A-3) for Rh-105, Sb-124, and Sb-125.

2-32 Ae.U l

.III U

II Sl, nfo,'tnr GI-LLI Rev. 18 1.05E-07 5.68E-07 1.70E-06 No Data 1.59E-09 No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data 1.05E-07 5.68E-07 1.70E-06 No Data 5.86E-10 1.36E-06 1.46E-07 No Data No Data No Data No Data No Data No Data 2.10E-07 1.03E-05 1.28E-08 No Data No Data No Data No Data No Data No Data No Data Lung 1.05E-07 5.68E-07 1.70E-06 No Data 3.53E-09 No Data No Data 1.06E-06 2.85E-06 No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data No Data 1.05E-07 5.68E-07 1.70E-06 2.17E-05 6.69E-07 1.40E-05 3.67E-06 1.09E-06 3.40E-05 1.51 E-05 4.02E-05 1.88E-06 1.74E-06 7.1 OE-06 9.70E-06 2.96E-09 5.79E-08 4.09E-1 3 No Data 4.16E-06 8.36E-1 9 2.33E-21 4.94E-05 2.19E-04 2.70E-05

VEGP ODCM Table 2-7 (contd).

Adult Ingestion Dose Factors Nuclide Bone I Liver T.Body Thyroid KidneyI Lung GI-LLI Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41 E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-1 1 No Data No Data No Data 1.48E-05 Y-93 2.68E-09 No Data 7.40E-1 1 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.1OE-05 Mo-99 No Data 4.31 E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.1OE-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-1 05 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31 E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41 E-05 Ag-i 1Dm 1.60E-07 1.48E-07 8.79E-08 No Data 2.91 E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71 E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.1 BE-08 7.65E-09 2.41 E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 2-33 Rev. 18 Rev. 18 2-33

VEGP ODCM "I A'-

A

• I..A.....

I\\

A4..1+ I 

Tneo attArQ I UI  ( IUUI ILu).

rUuIL II 3U'JI I

I Nuclide Bone 1 Liver T.Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71 E-05 1-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 1-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 1-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 1-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31 E-06 No Data 2.22E-06 1-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 1-135 4.43E-07 1.1 6E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31 E-06 Cs-134 6.22E-05 1.48E-04 1.21 E-04 No Data 4.79E-05 1.59E-05 2.59E-06 Cs-1 36 6.51 E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 2.11 E-06 Cs-1 38 5.52E-08 1.09E-07 5.40E-08 No Data 8.01 E-08 7.91 E-09 4.65E-1 3 Ba-139 9.70E-08 6.91 E-1 1 2.84E-09 No Data 6.46E-11 3.92E-1 1 1.72E-07 Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.OOE-26 La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 La-142 1.28E-10 5.82E-1 1 1.45E-11 No Data No Data No Data 4.25E-07 Ce-141 9.36E-09 6.33E-09 7.1BE-10 No Data 2.94E-09 No Data 2.42E-05 Ce-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21 E-07 No Data 1.65E-04 Pr-1 43 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01 E-11 I1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.03E-07 8.61 E-08 3.01 E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 2-34 Rev. 18 Rev. 18 2-34

VEGP ODCM Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.32E+00 1.32E+00 1.32E+00 1.32E+00 1.32E+00 1.32E+00 C-14 3.13E+04 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 Na-24 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 P-32 1.32E+06 8.22E+04 5.11 E+04 0.00 0.00 0.00 1.49E+05 Cr-51 0.00 0.00 1.27E+00 7.58E-01 2.79E-01 1.68E+00 3.19E+02 Mn-54 0.00 4.41 E+03 8.42E+02 0.00 1.31 E+03 0.00 1.35E+04 Mn-56 0.00 1.74E-01 3.08E-02 0.00 2.21 E-01 0.00 5.55E+00 Fe-55 6.86E+02 4.74E+02 1.11E+02 0.00 0.00 2.65E+02 2.72E+02 Fe-59 1.07E+03 2.51 E+03 9.61 E+02 0.00 0.00 7.01 E+02 8.36E+03 Co-58 0.00 9.59E+01 2.15E+02 0.00 0.00 0.00 1.94E+03 Co-60 0.00 2.78E+02 6.14E+02 0.00 0.00 0.00 5.23E+03 Ni-63 3.25E+04 2.25E+03 1.09E+03 0.00 0.00 0.00 4.70E+02 Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 2.75E+00 1.29E+00 0.00 6.94E+00 0.00 2.35E+02 Zn-65 2.32E+04 7.37E+04 3.33E+04 0.00 4.93E+04 0.00 4.64E+04 Zn-69 7.88E-07 1.51 E-06 1.05E-07 0.00 9.79E-07 0.00 2.26E-07 Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02 Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61E-18 Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Rb-86 0.00 9.75E+04 4.54E+04 0.00 0.00 0.00 1.92E+04 Rb-88 0.00 1.29E-22 6.82E-23 0.00 0.00 0.00 1.78E-33 Rb-89 0.00 1.61 E-26 1.14E-26 0.00 0.00 0.00 0.00 Sr-89 2.49E+04 0.00 7.16E+02 0.00 0.00 0.00 4.OOE+03 Sr-90 6.23E+05 0.00 1.53E+05 0.00 0.00 0.00 1.80E+04 Sr-91 7.25E+01 0.00 2.93E+00 0.00 0.00 0.00 3.45E+02 Sr-92 3.33E-01 0.00 1.44E-02 0.00 0.00 0.00 6.60E+00 Y-90 5.04E-01 0.00 1.35E-02 0.00 0.00 0.00 5.34E+03 Y-91m 1.04E-11 0.00 4.01E-13 0.00 0.00 0.00 3.04E-11 Y-91 9.77E+00 0.00 2.61 E-01 0.00 0.00 0.00 5.38E+03 Y-92 4.61 E-04 0.00 1.35E-05 0.00 0.00 0.00 8.07E+00 Y-93 3.19E-02 0.00 8.82E-04 0.00 0.00 0.00 1.01 E+03 Zr-95 5.47E-01 1.75E-01 1.19E-01 0.00 2.75E-01 0.00 5.56E+02 Zr-97 7.40E-03 1.49E-03 6.83E-04 0.00 2.26E-03 0.00 4.62E+02 Nb-95 8.09E+00 4.50E+00 2.42E+00 0.00 4.45E+00 0.00 2.73E+04 Mo-99 0.00 1.07E+02 2.04E+01 0.00 2.43E+02 0.00 2.49E+02 Tc-99m 5.70E-04 1.61 E-03 2.05E-02 0.00 2.44E-02 7.89E-04 9.53E-01 All values are in (mrem.mL)/(h.ltCi). They are calculated using equation (2.14), and data from Table 2-5, Table 2-6, and Table 2-7. When "No Data" is shown for a radionuclide-organ combination in Table 2-7, A,, factors in this table are presented as zero.

2-35 Rev. 18 "l',,* kit, OQ Rev. 18 2-35

VEGP ODCM Table 2-8 (contd).

Site-Related Ingestion Dose Factors, A1, Nuclide Bone Liver T. BodyI Thyroid Kidney Lung GI-LLI Tc-1 01 2.71 E-33 3.91 E-33 3.83E-32 0.00 7.03E-32 2.OOE-33 0.00 Ru-103 6.21 E+00 0.00 2.68E+00 0.00 2.37E+01 0.00 7.25E+02 Ru-105 8.79E-03 0.00 3.47E-03 0.00 1.14E-01 0.00 5.38E+00 Ru-106 9.42E+01 0.00 1.19E+01 0.00 1.82E+02 0.00 6.1OE+03 Rh-105 2.32E+00 1.69E+00 1.11E+00 0.00 7.15E+00 0.00 2.68E+02 Ag-110m 2.53E+00 2.34E+00 1.39E+00 0.00 4.61E+00 0.00 9.56E+02 Sb-124 1.36E+03 2.56E+01 5.37E+02 3.28E+00 0.00 1.05E+03 3.84E+04 Sb-125 1.09E+03 1.17E+01 2.19E+02 9.68E-01 0.00 1.14E+05 9.63E+03 Te-125m 2.56E+03 9.29E+02 3.43E+02 7.71 E+02 1.04E+04 0.00 1.02E+04 Te-127m 6.51 E+03 2.33E+03 7.93E+02 1.66E+03 2.64E+04 0.00 2.18E+04 Te-127 1.78E+01 6.40E+00 3.85E+00 1.32E+01 7.25E+01 0.00 1.41 E+03 Te-129m 1.09E+04 4.07E+03 1.73E+03 3.74E+03 4.55E+04 0.00 5.49E+04 Te-129 1.78E-05 6.68E-06 4.33E-06 1.36E-05 7.47E-05 0.00 1.34E-05 Te-131m 9.57E+02 4.68E+02 3.90E+02 7.42E+02 4.74E+03 0.00 4.65E+04 Te-131 8.64E-17 3.61E-17 2.73E-17 7.10E-17 3.78E-16 0.00 1.22E-17 Te-132 1.97E+03 1.27E+03 1.19E+03 1.41 E+03 1.23E+04 0.00 6.02E+04 1-130 7.60E+00 2.24E+01 8.85E+00 1.90E+03 3.50E+01 0.00 1.93E+01 1-131 1.73E+02 2.48E+02 1.42E+02 8.13E+04 4.25E+02 0.00 6.55E+01 1-132 5.27E-03 1.41 E-02 4.93E-03 4.93E-01 2.24E-02 0.00 2.65E-03 1-133 2.59E+01 4.51 E+01 1.37E+01 6.62E+03 7.86E+01 0.00 4.05E+01 1-134 2.18E-08 5.94E-08 2.12E-08 1.03E-06 9.44E-08 0.00 5.17E-1 1 1-135 1.31 E+00 3.44E+00 1.27E+00 2.27E+02 5.52E+00 0.00 3.89E+00 Cs-134 2.98E+05 7.10E+05 5.80E+05 0.00 2.30E+05 7.62E+04 1.24E+04 Cs-1 36 2.96E+04 1.17E+05 8.42E+04 0.00 6.51 E+04 8.92E+03 1.33E+04 Cs-137 3.82E+05 5.23E+05 3.43E+05 0.00 1.78E+05 5.90E+04 1.01 E+04 Cs-138 9.12E-12 1.80E-11 8.92E-12 0.00 1.32E-11 1.31E-12 7.68E-17 Ba-1 39 5.64E-06 4.02E-09 1.65E-07 0.00 3.76E-09 2.28E-09 1.OOE-05 Ba-140 3.74E+02 4.69E-01 2.45E+01 0.00 1.60E-01 2.69E-01 7.69E+02 Ba-141 8.47E-25 6.40E-28 2.86E-26 0.00 5.95E-28 3.63E-28 3.99E-34 Ba-142 0.00 0.00 0.00 0.00 0.00 0.00 0.00 La-140 1.10E-01 5.56E-02 1.47E-02 0.00 0.00 0.00 4.08E+03 La-142 2.19E-07 9.96E-08 2.48E-08 0.00 0.00 0.00 7.27E-04 Ce-141 1.15E-01 7.79E-02 8.84E-03 0.00 3.62E-02 0.00 2.98E+02 Ce-143 8.65E-03 6.39E+00 7.08E-04 0.00 2.81 E-03 0.00 2.39E+02 Ce-144 6.22E+00 2.60E+00 3.34E-01 0.00 1.54E+00 0.00 2.10E+03 Pr-1 43 6.10E-01 2.4.4E-01 3.02E-02 0.00, 1.41 E-01 0.00 2.67E+03 Pr-144 1.48E-28 6.14E-29 7.51 E-30 0.00 3.46E-29 0.00 2.13E-35.

Nd-147 4.11 E-01 4.75E-01 2.84E-02 0.00 2.78E-01 0.00 2.28E+0,3 W-187 1.47E+02 1.23E+021 4.31 E+01 0.00 0.00, 0.00 4.04E+04 ND-239

.1-2276-315E0 0.00 1.2-3 00 5.67E+02 2-36 Rev. 18 Rev. 18 2-36

VEGP ODCM 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2.5.1 Thirty-One Day Dose Projections In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to UNRESTRICTED AREAS of liquid effluents containing radioactive materials occurs or is expected.

Projected 31-day doses to individuals due to liquid effluents may be determined as follows:

D 'P (D,,)

x31+D,,

(2.17) where:

D*p

=

the projected dose to the total body or organ T, for the next 31 days of liquid releases.

D,

=

the cumulative dose to the total body or organ r, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

t

=

the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).

Da

=

the anticipated dose contribution to the total body or any organ t, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in addition to routine liquid effluents. If only routine liquid effluents are anticipated, D= may be set to zero.

2.5.2 Dose Proiections for Specific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology of Section 2.4, using sample analysis results for the source to be released, and parameter values expected to exist during the release period.

2-37 Rev. 18 Rev. 18 2-37

VEGP ODCM 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS The following symbolic terms are used in the presentation of liquid effluent calculations in the sub sections above.

Section of Term Definition Initial Use Ap=

the adjustment factor used in calculating the effluent monitor setpoint for liquid release pathway p: the ratio of the assured dilution to the required dilution [unitless].

2.3.2.2 ADF =

the assured dilution factor for a planned release [unitless].

2.3.2.2 AFp =

the dilution allocation factor for liquid release pathway p [unitless].

2.3.2.2 S=

the site-related adult ingestion dose commitment factor, for the total body or for any organ 'r, due to identified radionuclide i

[(mrem

• mL)/(h

• pICi)]. The values of A, are listed in Table 2-8.

2.4.1 B,=

the crop to soil concentration factor applicable to radionuclide i,

[(pCi/kg garden vegetation)/(pCi/kg soil)].

2.4.3 BF, =

the bioaccumulation factor for radionuclide i for freshwater fish

[(pCi/kg)/(pCVL)]. Values are listed in Table 2-6.

2.4.2 c =

the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line, prior to dilution and subsequent release [(CCVmL].

2.3.2.1 cp=

the calculated effluent radioactivity monitor setpoint for liquid release pathway p [CVmL].

2.3.2.2 Ca =

the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite sample [IiCi/mL].

2.3.2.2 CECL =

the Effluent Concentration Limit stated in 10 CFR 20, Appendix B, Table 2, Column 2 [JCVmL].

2.3.2.1 C=

the concentration of Fe-55 in the liquid waste as measured in the applicable composite sample [IiCi/mL].

2.3.2.2 Cg =

the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the applicable pre-release waste sample [gCi/mL].

2.3.2.2 Ci =

the measured concentration of radionuclide i in a sample of liquid effluent [gCi/mL].

2.3.2.2 2-38 Rev. 18 Rev. 18 2-38

VEGP ODCM Section of Term Definition Initial Use CU =

the average concentration of radionuclide i in undiluted liquid effluent during time period [CiCVmL].

2.4.1 Cir =

the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream [p.Ci/mL].

2.3.2.2 Cs =

the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste as measured in the applicable composite sample

[ICVmL].

2.3.2.2 Ct=

the concentration of H-3 in the liquid waste as measured in the applicable composite sample [gCi/mL].

2.3.2.2 CF, =

the concentration factor for radionuclide i in irrigated garden vegetation [(pCL/kg)/(pCi/L)].

2.4.2 D=

the dilution factor from the near field of the discharge structure to the potable water intake location [unitless].

2.4.2 DI =

the cumulative dose commitment to the total body or to any organ t, due to radioactivity in liquid effluents released during a given time period [mrem].

2.4.1 Dm =

the anticipated dose contribution to the total body or any organ 'r, due to any planned activities during the next 31-day period

[mrem].

2.5.1 D, =

the cumulative dose to the total body or organ T, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrem].

2.5.1 D~p =

the projected dose to the total body or organ T, for the next 31 days of liquid releases [mrem].

2.5.1 DFk =

the dose conversion factor for radionuclide i for adults, in organ 'r

[mrem/pCi]. Values are listed in Table 2-7.

2.4.2 ECL, =

the liquid Effluent Concentration Umit for radionuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 [Ci/mL].

2.3.2.2 f =

the effluent flowrate at the location of the radioactivity monitor

[gpm].

2.3.2.1 faP=

the anticipated actual discharge flowrate for a planned release from liquid release pathway p [gpm].

2.3.2.2 Rev. 18 2-39

VEGP ODCM Section of Term Definition Initial Use f=

the fraction of the year that garden vegetation is irrigated [unitless].

2.4.3 fmp =

the maximum permissible effluent discharge flowrate for release pathway p [gpm].

2.3.2.2 fr =

the effluent discharge flowrate of release pathway r [gpm].

2.3.2.2 ft =

the average undiluted liquid waste flowrate actually observed during the period of a liquid release [gpm].

2.4.1 F =

the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA [gpm].

2.3.2.1 Fd =

the entire assured dilution flowrate for the plant site during the release period [gpm].

2.3.2.2 Fd=

the dilution flowrate allocated to release pathway p [gpm].

2.3.2.2 F, =

the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA [unitless].

2.4.1 Ft =

the average dilution stream flowrate actually observed during the period of a liquid release [gpm].

2.4.1 I =

the average irrigation rate during the growing season [L/(m 2 -h)].

2.4.3 the water content of leafy garden vegetation edible parts [L/kg].

2.4.3 M =

the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage [unitless].

2.4.3 P =

the effective surface density of soil [kg/m2].

2.4.3 r =

the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.

2.4.3 RDF =

the required dilution factor: the minimum ratio by which liquid effluent must be diluted before reaching the UNRESTRICTED AREA, in order to ensure that the limits of Section 2.1.2 are not exceeded [unitless].

2.3.2.2 RDFr =

the RDF for a liquid release due only to its concentration of gamma-emitting radionuclides [unitless].

2.3.2.2 2-40 Rev. 18

VEGP ODCM Section of Term Definition Initial Use RDF =

the RDF for a liquid release due only to its concentration of non gamma-emitting radionuclides [unitless].

2.3.2.2 SF =

the safety factor selected to compensate for statistical fluctuations and errors of measurement [unitless].

2.3.2.2 t =

the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration.

2.5.1 tb=

the period of long-term buildup of activity in soil [h].

2.4.3 t' =

the period of leafy garden vegetation exposure during the growing season [h].

2.4.3 tf =

the transit time from release to receptor for fish consumption [h].

2.4.2 th =

the time between harvest of garden vegetation and human consumption [h].

2.4.3 t=

the transit time from release to receptor for potable water consumption [h].

2.4.2 TF =

the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate effluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2

[unitless]; the tolerance factor must not exceed a value of 10.

2.3.2.1 Uf=

the adult rate of fish consumption [kg/y].

2.4.2 UI=

the adult consumption rate for irrigated garden vegetation [kg/y].

2.4.2 Uw=

the adult drinking water consumption rate applicable to the plant site [L/y].

2.4.2 Yv=

the areal density (agricultural productivity) of leafy garden vegetation [kg/m2].

2.4.3 Z =

the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release [unitless].

2.4.1 At1 =

the length of time period 1, over which C11 and F1 are averaged for liquid releases [h].

2.4.1 2-41 Rev. 18

VEGP ODOM Definition the effective removal rate for activity deposited on crop leaves

[hl].

the decay constant for radionuclide i [h'].

the rate constant for removal of activity from plant leaves by weathering [h'].

242 Rev. 18 Term

%Ei --

k. =

XW=

Section of Initial Use 2.4.3 2.4.2 2.4.3 Rev. 18 2-42

VEGP ODCM CHAPTER 3 GASEOUS EFFLUENTS 3.1 LIMITS OF OPERATION The following Limits of Operation implement requirements established by Technical Specifications Section 5.0. Terms printed in all capital letters are defined in Chapter 10.

3.1.1 Gaseous Effluent Monitoring Instrumentation Control In accordance with Technical Specification 5.5.4.a, the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Section 3.1.2.a are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with Section 3.3.

3.1.1.1 Applicability These limits apply as shown in Table 3-1.

3.1.1.2 Actions With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of Section 3.1.2.a are met.

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3-1. Restore the inoperable instrumentation to operable status within 30 days, or if unsuccessful, explain in the next Radioactive Effluent Release Report, per Technical Specification 5.6.3, why this inoperability was not corrected in a timely manner.

This control does not affect shutdown requirements or MODE changes.

3.1.1.3 Surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 3-2.

3-1 Rev. 18 Rev. 18 3-1

VEGP ODCM 3.1.1.4 Basis The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 3.3 to ensure that the alarm/trip will occur prior to exceeding the limits of Section 3.1.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

Rev. 18 3-2

VEGP ODCM Table 3-1.

Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirements Minimum Instrument Channels OPERABLE Applicability ACTION

1. GASEOUS RADWASTE TREATMENT SYSTEM (Common)

a. Noble Gas Activity Monitor, with Alarm and Automatic Termination of Release (ARE-0014) 1 During releasesa 45

b. Effluent System Flowrate Measuring Device (AFT-0014) 1 During releasesa 46

2. Turbine Building Vent (Each Unit)

a. Noble Gas Activity Monitor (RE-12839C) 1 During releasesa 47

b. Iodine and Particulate Samplers (RE-12839A & B) 1 During releasesa 51

c. Flowrate Monitor (FT-12839 or FIS-1 2 8 6 2 )b 1

During releasesa 46

d. Sampler Flowrate Monitor (Fl-1 3211) 1 During releasesa 46

3. Plant Vent (Each Unit)

a. Noble Gas Activity Monitor (RE-12442C or RE-12444C) 1 At all times 47,48

b. Iodine Sampler/Monitor (RE-12442B or RE-12444B) 1 At all times 51

c. Particulate Sampler/Monitor (RE-12442A or RE-12444A) 1 At all times 51

d. Flowrate Monitor (FT-12442 or 12835) 1 At all times 46

e. Sampler Flowrate Monitor (FI-12442 or FI-12444) 1 At all times 46

4. Radwaste Processing Facility Vent (Common)

a. Particulate Monitor (ARE-1 6980) 1 During Releases 51

a.

b.

"During releases" means 'During radioactive releases via this pathway."

During emergency filtration.

Rev. 19 3-3

VEGP ODCM Table 3-1 (contd).

Notation for Table 3 ACTION Statements ACTION 45 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

a.

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

b.

At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and verify the release rate calculations.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 47 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend containment purging of radioactive effluents via this pathway.

ACTION 49 - (Not Used)

ACTION 50 - (Not Used)

ACTION 51 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment.

3-4 Rev. 18 Rev. 18 3-4

VEGP ODCM Table 3-2.

Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Surveillance Requirements CHANNEL CHANNEL OPERA Instrument CHANNEL SOURCE CALIBRA-TIONAL CHECK CHECK TION TEST MODESc

1. GASEOUS RADWASTE TREATMENT SYSTEM (Common)

a. Noble Gas Activity Monitor, with Alarm and Automatic Termination of Release During (ARE-0014)

P P

Rb Qa(l)

Release

b. Effluent System Flowrate Measuring Device During (AFT-0014)

P NA R

NA Release

2. Turbine Building Vent (Each Unit)

a. Noble Gas Activity Monitor b

During (RE-12839C)

D M

Rb Qa(2)

Release

b. Iodine and Particulate During Samplers (RE-12839A&B)

Wd NA NA NA Release

c. Flowrate Monitor During (FT-12839 or FIS-12862)

D NA R

NA Release

d. Sampler Flowrate Monitor During (FI-13211)

D NA R

Q Release

3. Plant Vent (Each Unit)

a. Noble Gas Activity Monitor (RE-12442C or b

RE-12444C)

D M

RQa()

All

b. Particulate and Iodine Monitors (RE-12442A&B)

Wd NA R

Qa(2)

All

c. Particulate and Iodine Samplers (RE-12444A&B)

Wd NA NA NA All

d. Flowrate Monitor (FT-12442 or 12835)

D NA R

NA All

e. Sampler Flowrate Monitor (FI-12442 or Fl-12444)

D NA R

Q All

4. Radwaste Processing Facility Vent (Common)

a. Particulate Monitor Wd ADuring (ARE-16980)

Release 3-5 Rev. 19 Rev. 19 3-5

VEGP ODCM Table 3-2 (contd).

Notation for Table 3-2

a.

In addition to the basic functions of a CHANNEL OPERATIONAL TEST (Section 10.2):

(1)

The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room annunciation occurs (for item a. below only); and control room CRT indication occurs (if any of the following conditions exist):

(a)

Instrument indicates measured levels above the alarm/trip setpoint; (b)

Instrument indicates an "Equipment Trouble" alarm; (c)

Instrument indicates a "Low" alarm; or (d)

Instrument indicates channel "Deactivated."

(2)

The CHANNEL OPERATIONAL TEST shall also demonstrate that control room annunciation occurs (for item a. below only); and that control room CRT indication occurs (if any of the following conditions exist):

(a)

Instrument indicates measured levels above the alarm/trip setpoint; (b)

Instrument indicates an "Equipment Trouble" alarm; (c)

Instrument indicates a "Low" alarm; or (d)

Instrument indicates channel "Deactivated." ("Loss of counts" for ARE-1 6980 only)

b.

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

c.

MODES in which surveillance is required. "All" means "At all times." "During release" means "During radioactive release via this pathway."

d.

The channel check shall consist of visually verifying that the collection device (i.e.,

particulate filter or charcoal cartridge, etc.) is in place for sampling.

e.

The CHANNEL CALIBRATION verifies proper operation of the CHANNEL OPERATIONAL TEST requirements described in Notation a(2) above.

3-6 Rev. 19 Rev. 19 3-6

VEGP ODCM 3.1.2 Gaseous Effluent Dose Rate Control In accordance with Technical Specifications 5.5.4.c and 5.5.4.g, the licensee shall conduct operations so that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY are limited as follows:

a.

For noble gases: Less than or equal to a dose rate of 500 mrem/y to the total body and less than or equal to a dose rate of 3000 mrem/y to the skin, and

b.

For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/y to any organ.

3.1.2.1 Applicability This limit applies at all times.

3.1.2.2 Actions With a dose rate due to radioactive material released in gaseous effluents exceeding the limit stated in Section 3.1.2, immediately decrease the release rate to within the stated limit.

These limits do not affect shutdown requirements or MODE changes.

3.1.2.3 Surveillance Requirements The dose rates due to radioactive materials in areas at or beyond the SITE BOUNDARY due to releases of gaseous effluents shall be determined to be within the above limits, in accordance with the methods and procedures in Section 3.4.1, by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3-3.

3.1.2.4 Basis This control is provided to ensure that gaseous effluent dose rates will be maintained within the limits that historically have provided reasonable assurance that radioactive material discharged in gaseous effluents will not result in a dose to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.

The dose rate limit for Iodine-1 31, lodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days specifically applies to dose rates to a child via the inhalation pathway.

This control applies to the release of gaseous effluents from all reactors at the site.

3-7 Rev. 18 Rev. 18 3-7

VEGP ODCM Table 3-3.

Radioactive Gaseous Waste Sampling and Analysis Program Sampling and Analysis Requirementsa MINIMUM DETECTABLE Minimum CONCENTRA Gaseous Sampling Analysis Type of Activity TION (MDC)

Release Type FREQUENCY FREQUENCY Analysis (g*Ci/mL)

Waste Gas P

P Noble Gas 1 E-4 Decay Tank Each Tank Grab Each Tank PRINCIPAL (Common)

Sample GAMMA EMITTERS pC Noble Gas 1 E-4 Containment pc Each Purge PRINCIPAL Purge Each Purge GAMMA EMITTERS 24" or 14" (Each Unit)

Grab Sample M

H-3 (Oxide) 1 E-6 Noble Gas 1 E-4 Plant Vent Mc'd'f Mc PRINCIPAL (Each Unit)

Grab Sample GAMMA EMITTERS H-3 (Oxide) 1 E-6 Condenser Air Noble Gas 1 E-4 Ejtonesr Air PRINCIPAL Ejector &

M MGMAEITR Steam Packing Grab Sample M

GAMMA EMITTERS Exhaust (Each Unit)b H-3 (Oxide) 1 E-6 We 1-131 1 E-12 Charcoal or Silver Zeolite Sample We Particulate 1 E-11 CONTINUOUSg Particulate PRINCIPAL Sample GAMMA EMITTERS Plant Vent, M

Gross Alpha 1 E-11 Condenser Air COMPOSITE Ejector &

CONTINUOUSg Particulate Steam Packing Sample Exhaust (1Each Unit) a Sr-89, Sr-90 1 E-11 COMPOSITE Particulate Sample Radwaste Wh Particulate 1 E-11 Processing Cg Particulate PRINCIPAL Facility Vent Sample GAMMA EMITTERS (Common) 3-8 Rev. 19 Rev. 19 3-8

VEGP ODCM Table 3-3 (contd).

Notation for Table 3-3

a.

Terms printed in all capital letters are defined in Chapter 10.

b.

The turbine building vent is the release point for the condenser air ejector and steam packing exhaust. All sampling and analyses may be omitted for this vent, provided the absence of a primary to secondary leak has been demonstrated, that is, if the gamma activity in the secondary water does not exceed background by more than 20%.

c.

Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one-hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

d.

Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling cavity is flooded.

e.

Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in one hour, and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding MDC may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

f.

Tritium grab samples shall be taken at least once per 7 days from the Unit 1 plant vent, whenever spent fuel is in the spent fuel pool (Unit 1 plant vent only).

g.

The ratio of the sample flowrate to the sampled stream flowrate shall be known for the time period covered by each dose or dose rate calculation made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.

h.

Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or removal of sampler).

3-9 Rev. 19 Rev. 19 3-9

VEGP ODCM 3.1.3 Gaseous Effluent Air Dose Control In accordance with Technical Specifications 5.5.4.e and 5.5.4.h, the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following:

a.

During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b.

During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

3.1.3.1 Applicability This limit applies at all times.

3.1.3.2 Actions With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which identifies the cause(s) for exceeding the limit(s); defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3.

This control does not affect shutdown requirements or MODE changes.

3.1.3.3 Surveillance Requirements Cumulative air dose contributions from noble gas radionuclides released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.2 at least once per 31 days.

3.1.3.4 Basis This control is provided to implement the requirements of Sections l1.B, III.A and IV.A of Appendix I, 10 CFR Part 50. Section 3.1.3 implements the guides set forth in Section lI.B of Appendix I.

The ACTION statements in Section 3.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I, assuring that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III.A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOUNDARY are based upon the historical annual average atmospheric conditions.

3-10 Rev. 18 Rev. 18 3-10

VEGP ODCM 3.1.4 Control on Gaseous Effluent Dose to a Member of the Public In accordance with Technical Specifications 5.5.4.e and 5.5.4.i, the dose to a MEMBER OF THE PUBLIC from 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following:

a.

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

b.

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

3.1.4.1 Applicability This limit applies at all times.

3.1.4.2 Actions With the calculated dose from the release of 1-131,1-133, tritium, or radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which identifies the cause(s) for exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radioiodines and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents; and defines proposed corrective actions to assure that subsequent releases will be in compliance with the limits stated in Section 3.1.4.

This control does not affect shutdown requirements or MODE changes.

3.1.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days.

3.1.4.4 Basis This control is provided to implement the requirements of Section II.C, III.A and IV.A of Appendix 1, 10 CFR Part 50. The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section 3.1.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The calculational methods specified in the Surveillance Requirements of Section 3.1.4.3 implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The calculational methods in Section 3.4.3 for calculating the doses due to the actual releases of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses Rev. 18 3-11

VEGP ODCM based upon the historical annual average atmospheric conditions. The release specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were:

1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3.1.5 Gaseous Radwaste Treatment System Control In accordance with Technical Specification 5.5.4.f, the GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous wastes prior to their discharge when the projected doses in 31 days due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY would exceed 0.2 mrad to air from gamma radiation, 0.4 mrad to air from beta radiation, or 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

3.1.5.1 Applicability These limits apply at all times.

3.1.5.2 Actions With gaseous waste being discharged without treatment and in excess of the limits in Section 3.1.5, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which includes the following information:

a.

Identification of any inoperable equipment or subsystem and the reason for inoperability,

b.

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

c.

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

This control does not affect shutdown requirements or MODE changes.

3.1.5.3 Surveillance Requirements Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days, in accordance with Section 3.5.1, when the GASEOUS WASTE PROCESSING SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

The GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE:

Rev. 18 3-12

VEGP ODCM by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS WASTE PROCESSING SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREATMENT SYSTEM).

3.1.5.4 Basis The OPERABILITY of the GASEOUS WASTE PROCESSING SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

3.1.6 Maior Chanaes to Gaseous Radioactive Waste Treatment Systems Licensee initiated MAJOR CHANGES TO GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEMS:

a.

Shall be reported to the Nuclear Regulatory Commission in the Radioactive Effluents Release Report for the period in which the change was implemented. The discussion of each change shall contain the information described in Section 7.2.2.7.

b.

Shall become effective upon review and approval by the General Manager -

Nuclear Plant.

3-13 Rev. 18 Rev. 18 3-13

VEGP ODCM 3.2 GASEOUS WASTE PROCESSING SYSTEM At Plant Vogtle, there are six potential points where radioactivity may be released to the atmosphere in gaseous discharges. These six potential release pathways are the Unit 1 and Unit 2 Plant Vents; the Unit 1 and Unit 2 Turbine Building Vents; the Radwaste Processing Facility Vent; and the Dry Active Waste Processing Building Vent. However, the Turbine Building Vents are not normal release pathways unless a primary-to-secondary leak exists. The Radwaste Processing Facility Vent is not a normal release pathway unless a spill occurs. The figures on the following pages give schematic diagrams of the Gaseous Waste Treatment System and the Ventilation Exhaust Treatment Systems (Reference 11).

The Unit 1 Plant Vent release pathway includes two release sources that are common to the two units: ventilation air from the Fuel Handling Building, and discharges from the GASEOUS WASTE PROCESSING SYSTEM. Otherwise, discharges from the two reactor units are separated. Reactor Containment Building ventilation releases are through the respective plant vents. The Turbine Building Vent serves as the discharge point for both the condenser air ejector and the steam packing exhauster system. The Radwaste Processing Facility Vent includes sources from the Radwaste Processing Facility Process area.

Releases from the two Turbine Building Vents, the Radwaste Processing Facility Vent, and the Dry Active Waste Processing Building Vent are considered to be ground-level releases, whereas releases from the two Plant Vents are considered mixed-mode releases. Chapter 8 discusses the calculation of atmospheric dispersion parameters using the ground-level and mixed-mode (i.e.,

split-wake) models. All six potential release pathways are considered to be continuous (as opposed to batch) in nature.

3-14 Rev. 19 Rev. 19 3-14

VEGP ODCM To Chemical Volume Control Tank To Waste Gas Decay Tank Header Unit 2 Waste Gas Volume Control Tank Purge Recycle Evaporator Vent Condenser Waste Evaporator Vent Condenser Recycle Holdup Tank Eductor Reactor Coolant Drain Tank

'Dotted line operational between 20 and 100 psig NOTE:

This is typical of both units. However, Unit 2 GWPS releases via Unit I plant vent.

Figure 3-1.

Schematic Diagram of the Gaseous Radwaste Treatment System 3-15 Rev. 18 I I I

3-15 Rev. 18

VEGP ODOM Plant Vent Radioactivity Monitor 1RE12442A,B,C Radioactivity Reactor Containment Monitor ARE-0014 Building From Waste Gas Processing Area and System HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil ME - Moisture Eliminator

• Prior to treatment by the Fuel Handling Building Ventilation Exhaust Treatment System, Exhaust from Unit I Spent Fuel Pool Area is monitored by ARE2532B and ARE2533B; exhaust from Unit 2 Spent Fuel Pool Area is monitored by ARE2532A and ARE2533A.

Figure 3-2. Schematic Diagram of the Unit 1 Plant Vent Release Pathway 3-16 Rev. 18 Rev. 18 3-16

VEGP ODCM Plant Vent Radioactivity Monitor 2RE12442A,B,C Auxiliary Building Radioactivity Monitor 2RE2565A,B,C Reactor Containment Building HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil ME - Moisture Eliminator Figure 3-3. Schematic Diagram of the Unit 2 Plant Vent Release Pathway Rev. 18 3-17

VEGP ODOM Turbine Building Vent t

Radioactivity Monitor 1(2)RE12839A,B,C NO NC LmNO NC Steam Jet Air Ejector Steam Packing Exhauster HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil DE - Demister NO - Normally Open NC - Normally Closed NOTE: This is typical of both units.

Figure 3-4.

Schematic Diagram of the Turbine Building Vent Release Pathway (Typical of Both Units) 3-18 Rev. 18

VEGP ODCM Dry Active Waste Processing Building Vent Radioactivity Monitor ARE 13256 Rooms Trash Compactor Figure 3-5.

Schematic Diagram of the Dry Active Waste Processing Building Ventilation Release Pathway 3-19 Rev. 18 Rev. 18 3-19

VEGP ODCM Figure 3-6.

Schematic Diagram of the Radwaste Processing Facility Ventilation Release Pathway Rev. 19 3-19a

VEGP ODCM 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3.3.1 General Provisions Regarding Noble Gas Monitor Setpoints Noble gas radioactivity monitor setpoints calculated in accordance with the methodology presented in this section are intended to ensure that the limits of Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint.

If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be calculated in accordance with the methods presented below, based on an assumed concentration of Kr-88 that leads to a practical setpoint. A practical setpoint in this context is one which prevents spurious alarms, and yet produces an alarm should a significant inadvertent release occur.

Section 3.1.1 establishes the requirements for gaseous effluent monitoring instrumentation, and Section 3.2 describes the VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS WASTE PROCESSING SYSTEM. From those Sections, it can be seen that certain monitors are located on final release pathways, that is, streams that are being monitored immediately before being discharged from the plant; the setpoint methodology for these monitors is presented in Section 3.3.2. Other monitors are located on source streams, that is, streams that merge with other streams prior to passing a final monitor and being discharged; the setpoint methodology for these monitors is presented in Section 3.3.3. Table 3-4 identifies which of these setpoint methodologies applies to each monitor. Some additional monitors with special setpoint requirements are discussed in Section 3.3.5.

As established in Section 3.1.1, gaseous effluent monitor setpoints are required only for the noble gas monitors on certain potential release streams: the two Plant Vents, the two Turbine Building Vents, and the GASEOUS WASTE PROCESSING SYSTEM discharge. However, because of the potential significance of releases from other sources, Section 3.3 discusses setpoint methodologies for certain additional monitors, as well.

Rev. 18 3-20

VEGP ODCM Table 3-4.

Applicability of Gaseous Monitor Setpoint Methodologies Final Release Pathways with no Monitored Source Streams Setpoint Method:

Section 3.3.2 Release Elevation:

Ground-level Unit 1 or Unit 2 Turbine Buildina Vent Monitor:

1 RE-12839C/2RE-12839C Maximum Flowrate:

900 cfm (4.25 E+05 mL/s)

Dry Active Waste Building Vent Monitor:

ARE-1 3256 Maximum Flowrate:

2,200 cfm (1.04 E+06 ml~s)

Final Release Pathways with One or More Monitored Source Streams Release Elevation: Mixed-Mode Unit 1 Plant Vent Monitors:

1 RE-12442C, 1 RE-12444C Maximum Flowrate:

187,000 cfm (8.83 E+07 mL/s)

Setpoint Method:

Section 3.3.2 Release Type:

CONTINUOUS Source Stream: Unit 1 Reactor Containment Puraqe Monitor:

1 RE-2565C Maximum Flowrate:

release-dependent Setpoint Method:

Section 3.3.3 Release Type:

BATCH Source Stream: Gaseous Waste Treatment System Monitor:

ARE-0014 Maximum Flowrate:

release-dependent Setpoint Method:

Section 3.3.3 Release Type:

BATCH Unit 2 Plant Vent Monitors:

2RE-12442C, 2RE-12444C Maximum Flowrate:

112,500 cfm (5.31 E+07 mL/s)

Setpoint Method:

Section 3.3.2 Release Type:

CONTINUOUS Source Stream: Unit 2 Reactor Containment Purqe Monitor:

2RE-2565C Maximum Flowrate:

release-dependent Setpoint Method:

Section 3.3.3 Release Type:

BATCH (XT)7, Values for Use in Setpoint Calculations Ground-Level Releases:

2.55 x 10 s/im3 [NE Sector]

Mixed-Mode Releases:

4.62 x 1 0V s/m3 [NE Sector]

Maximum flowrate values are from Reference 11, Table 11.5.2-1 and Table 11.5.5-1.

3-21 Rev. 18

VEGP ODCM 3.3.2 Setpoint for the Final Noble Gas Monitor on Each Release Pathway 3.3.2.1 Overview of Method Gaseous effluent radioactivity monitors are intended to alarm prior to exceeding the limits of Section 3.1.2.a. Therefore, their alarm setpoints are established to ensure compliance with the following equation:

c=the lesser o AGSFXR (3.1) where:

c

=

the setpoint, in gCi/mL, of the radioactivity monitor measuring the concen tration of radioactivity in the effluent line prior to release. The setpoint represents a concentration which, if exceeded, could result in dose rates exceeding the limits of Section 3.1.2.a at or beyond the SITE BOUNDARY.

AG

=

an administrative allocation factor applied to divide the release limit among all the gaseous release pathways at the site.

SF

=

the safety factor selected to compensate for statistical fluctuations and errors of measurement.

X

=

the noble gas concentration for the release under consideration.

Rt

=

the ratio of the dose rate limit for the total body, 500 mrem/y, to the dose rate to the total body for the conditions of the release under consideration.

Rk

=

the ratio of the dose rate limit for the skin, 3000 mrem/y, to the dose rate to the skin for the conditions of the release under consideration.

Equation (3.1) shows the relationships of the critical parameters that determine the setpoint.

However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology.

The basic setpoint method presented below is applicable to the radioactivity monitor nearest the point of release for the release pathway. For monitors measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section 3.3.3 must be applied.

3.3.2.2 Setpoint Calculation Steps Sten 1:

Determine the concentration, X4, of each noble gas radionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analysis requirements of Section 3.1.2. Then sum these concentrations to determine the total noble gas concentration, 14.

3-22 Rev. 18

VEGP ODCM Step 2:

Determine Rt, the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the release under consideration, as follows:

Rt =

500 (3.2)

(X/Q) vb Y. [K,

• Q..]

where:

500

=

the dose rate limit for the total body, 500 mrem/y.

(X/Q)*

=

the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v. Table 3-4 includes an indication of what release elevation is applicable to each release pathway; release elevation determines the appropriate value of (X/Q),

K,

=

the total-body dose factor due to gamma emissions from noble gas radio nuclide i, in (mrem/y)I(g.Cilm3), from Table 3-5.

Qi,

=

the release rate of noble gas radionuclide i from the release pathway under consideration, in jiCi/s, calculated as the product of X4, and fa,, where:

X

=

the concentration of noble gas radionuclide i for the particular release, in lCi/mL.

f=

the maximum anticipated flowrate for release pathway v during the period of the release under consideration, in mL/s.

Stec) 3:

Determine Rk, the ratio of the dose rate limit for the skin, 3000 mrem/y, to the skin dose rate due to noble gases detected in the release under consideration, as follows:

3000 (3.3)

R*

I(-)vb

.[(Li,+ I.IM,).-QN, where:

3000

= the dose rate limit for the skin, 3000 mrem/y.

L

=

the skin dose factor due to beta emissions from noble gas radionuclide i, in (mrem/y)/(gCVCm3), from Table 3-5.

M1

=

the air dose factor due to gamma emissions from noble gas radionuclide i, in (mrad/y)/(lgCi/m3), from Table 3-5.

1.1

=

the factor to convert air dose in mrad to skin dose in mrem.

All other terms were defined previously.

3-23 Rev. 18

VEGP ODCM SteD 4:

Determine the maximum noble gas radioactivity monitor setpoint concentration.

Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 3.1.2.a will not be exceeded.

Because the radioactivity monitor responds primarily to radiation from noble gas radionuclides, the monitor setpoint c,, (in giCi/mL) is based on the concentration of all noble gases in the waste stream, as follows:

where:

Cnv

=

the calculated setpoint, in gCGi/mL, for the noble gas monitor serving gaseous release pathway v.

F AGv "SF" -

XXv Rt enV =the lesser of i

(3.4)

LAGV *SF. X Xiv Rk i

AGV

=

the administrative allocation factor for gaseous release pathway v, applied to divide the release limit among all the gaseous release pathways at the site.

The allocation factor may be assigned any value between 0 and 1, under the condition that the sum of the allocation factors for all simultaneously-active final release pathways at the entire plant site does not exceed 1. Alternative methods for determination of AG, are presented in Section 3.3.4.

SF

=

the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1.

A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired.

X

=

the measured concentration of noble gas radionuclide i in gaseous stream v, as defined in Step 1, in liCi/mL.

The values of Rt and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above.

Steg 5:

Determine whether the release is permissible, as follows:

If c,, > Z, the release is permissible. However, if cm, is within about 10 percent of,_Xv,, it may be impractical to use this value of cr,. This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.

3-24 Rev. 18 Rev. 18 3-24

VEGP ODCM If c,, < _X,, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

3.3.2.3 Use of the Calculated Setpoint The setpoint calculated above is in the units iCi/mL. The monitor actually measures a count rate, subtracts a predetermined background count rate, and multiplies by a calibration factor to convert from count rate to p.Ci/mL.

Initial calibration by the manufacturer and Georgia Power Company of the gaseous effluent monitors specified in Section 3.1.1 utilized at least one NIST-traceable gaseous radionuclide source in the exact geometry of each production monitor. The point and gaseous sources used covered the beta particle end point energy range from 0.293 MeV to at least 1.488 MeV. The calibration factor is a function of the radionuclide mix in the gas to be released, and normally will be calculated for the monitor based on the results of the sample results from the laboratory gamma-ray spectrometer system. The mix-dependent calibration factor will be used as the gain factor in the PERMS monitor, or used to modify the calculated base monitor setpoint so that the default calibration factor in the PERMS monitor can be left unchanged.

Notwithstanding the initial calibration, monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to stream concentrations measured by sample analysis.

In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. Contributions to the monitor background may include any or all of the following factors: ambient background radiation, plant-related radiation levels at the monitor location (which may change between shutdown and power conditions), and internal background due to contamination of the monitor's sample chamber.

3.3.3 Setpoints for Noble Gas Monitors on Effluent Source Streams Table 3-4 lists certain gaseous release pathways as being source streams. As may be seen in the figures of Section 3.2, these are streams that merge with other streams, prior to passing a final radioactivity monitor and being released. Unlike the final monitors, the source stream monitors measure radioactivity in effluent streams for which flow can be terminated; therefore, the source stream monitors have control logic to terminate the source stream release at the alarm setpoint.

3.3.3.1 Setpoint of the Monitor on the Source Stream Ste_ 1:

Determine the concentration Xs of each noble gas radionuclide i in source stream s (in gCi/mL) according to the results of its required sample analyses [see Section 3.1.2].

SteD 2:

Determine rt, the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and the maximum anticipated source stream Rev. 18 3-25

VEGP ODCM flowrate fas in equation (3.2) to determine the total body dose rate for the source stream, substituting rt for Rt.

The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the source stream is the same as the (X/Q)* that applies to the respective merged stream. This is because the (X/Q) value is determined by the meteorology of the plant site and the physical attributes of the release point, and is unaffected by whether or not a given source stream is operating.

Steo 3:

Determine rk, the ratio of the dose rate limit for the skin, 3000 mrem/y, to the skin dose rate due to noble gases detected in the source stream under consideration.

Use the Xis values and the maximum anticipated source stream flow rate fas in equation (3.3) to determine the skin dose rate for the source stream, substituting rk for Rk.

Steg 4:

Determine the maximum noble gas radioactivity monitor setpoint concentration, as follows:

fAGS.SF. X

.,ri cs =the lesser of i

(3.5)

LAGs *SF.

XX, *rk where:

Cns

=

the calculated setpoint (in gCVmL) for the noble gas monitor serving gaseous source stream s.

AG,

=

the administrative allocation factor applied to gaseous source stream s. For a given final release point v, the sum of all the AG, values for source streams contributing to the final release point must not exceed the release point's allocation factor Agv.

Xis

=

the measured concentration of noble gas radionuclide i in gaseous source stream s, as defined in Step 1, in gCVmL.

The values of rt and rk to be used in the calculation are those which were determined in Steps 2 and 3 above. The safety factor, SF, was defined previously.

Step 5:

Determine whether the release is permissible, as follows:

If c,s >

2.

Xi, the release is permissible. However, if crs is within about 10 percent of Yis, it may be impractical to use this value of c,s. This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.

Rev. 18 3-26

VEGP ODCM If cn, <.Xj, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

3.3.3.2 Effect on the Setpoint of the Monitor on the Merged Stream Before beginning a release from a monitored source stream, a setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the previously-determined maximum allowable setpoint for the downstream final monitor on the merged stream must be redetermined. This is accomplished by repeating the steps of Section 3.3.2, with the following modifications.

Modification 1:

The new maximum anticipated flowrate of the merged stream is the sum of the old merged stream maximum flowrate, and the maximum flowrate of the source stream being considered for release.

(

= (fs )od + fa (3.6)

Modification 2:

The new concentration of noble gas radionuclide i in the merged stream includes both the contribution of the merged stream without the source stream, andthe source stream being considered for release.

(Xv)new (fav )old (Xi' )old + f.= + Xis (3.7) 3.3.4 Determination of Allocation Factors, AG When simultaneous gaseous releases are conducted, an administrative allocation factor must be applied to divide the release limit among the active gaseous release pathways. This is to assure that the dose rate limit for areas at and beyond the SITE BOUNDARY (see Section 3.1.2) will not be exceeded by simultaneous releases. The allocation factor for any pathway may be assigned any value between 0 and 1, under the following two conditions:

1.

The sum of the allocation factors for all simultaneously-active final release paths at the plant site may not exceed 1.

2.

The sum of the allocation factors for all simultaneously-active source streams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.

Any of the following three methods may be used to assign the allocation factors to the active gaseous release pathways:

1.

For ease of implementation, AGv may be equal for all release pathways:

AG,=

(3.8) 3-27 Rev. 18 Rev. 18 3-27

VEGP ODCM where:

N

=

the number of simultaneously active gaseous release pathways.

2.

AG, for a given release pathway may be selected based on an estimate of the portion of the total SITE BOUNDARY dose rate (from all simultaneous releases) that is contributed by the release pathway. During periods when a given building or release pathway is not subject to gaseous radioactive releases, it may be assigned an allocation factor of zero.

3.

AGv for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows:

ýXI/Q)Vb X(KiQ,)

AGv = N (3.9) where:

(X/Q),

=

the annual average SITE BOUNDARY relative concentration applicable to the gaseous release pathway v for which the allocation factor is being determined, in s/m3.

=

the total-body dose factor due to gamma emissions from noble gas radio nuclide i, in (mrem/y)/(p.lCVm 3), from Table 3-5.

Qi,

=

the release rate of noble gas radionuclide i from release pathway v, in jiCi/s, calculated as the product of X, and fay, where:

X

=

the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the allocation factor is being determined, in p!Ci/mL.

fay

=

the discharge flowrate applicable to gaseous release pathway v for which the allocation factor is being determined, in mL/s.

(X/Q)rb

=

the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r, in s/m3.

oil

=

the release rate of noble gas radionuclide i applicable to active release pathway r, in lCi/s, calculated as the product of Xi, and far, where:

X

=

the concentration of noble gas radionuclide i applicable to active gaseous release pathway r, in ViCi/mL.

far

=

the discharge flowrate applicable to active gaseous release pathway r, in mL/s.

Rev. 18 3-28

VEGP ODCM N

=

the number of simultaneously active gaseous release pathways (including pathway v that is of interest).

NOTE:

Although equations (3.8) and (3.9) are written to illustrate the assignment of the allocation factors for final release pathways, they may also be used to assign allocation factors to the source streams that merge into a given final release pathway.

3.3.5 Setpoints for Noble Gas Monitors with Special Requirements At present, VEGP has no noble gas monitors for which setpoint methodologies are to be presented in the ODCM, and that require methods other than those in Section 3.3.2 or Section 3.3.3.

3.3.6 Setpoints for Particulate and Iodine Monitors In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the ODCM establish setpoint calculation methods for particulate and iodine monitors. Therefore, the following is provided for information only: Initial setpoints for the particulate channels of effluent monitors RE-1 2442, RE-2565, and ARE-1 3256 were determined as described in Reference 13.

3-29 Rev. 18 Rev. 18 3-29

VEGP ODCM 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3.4.1 Dose Rates at and Beyond the Site Boundary Because the dose rate limits for areas at and beyond the SITE specified in Section 3.1.2 are site limits applicable at any instant in time, the summations extend over all simultaneously active gaseous final release pathways at the plant site. Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (X/Q)* value for each.

3.4.1.1 Dose Rates Due to Noble Gases For the purpose of implementing the controls of Section 3.1.2.a, the dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:

For total body dose rates:

DR, X

{(/ Q)V X[KiQjI (3.10)

For skin dose rates:

DRk

{xI-Q)Vb XI(Li + 1.lM,)Qiv}

(3.11) where:

DRt

=

the total body dose rate at the time of the release, in mrem/y.

DRk

=

the skin dose rate at the time of the release, in mrem/y.

Qj,

=

the release rate of noble gas radionuclide i, in jiCi/s, equal to the product of ft, and X,, where:

ft

=

the actual average flowrate for release pathway v during the period of the release, in mL/s.

All other terms were defined previously.

3.4.1.2 Dose Rates Due to Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form with Half-Lives Greater than 8 Days For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to lodine-131, Iodine-1 33, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:

DR, ={(X/Q),vb

[Pi. Q i}

(3.12) 3-30 Rev. 18 Rev. 18 3-30

VEGP ODCM where:

DR,

=

the dose rate to organ o at the time of the release, in mrem/y.

Pi.

=

the site-specific dose factor for radionuclide i and organ o, in (mrem/y)/(gCi/m 3). Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation pathway, the values of Pi, may be obtained from Table 3-9, "Rapj for Inhalation Pathway, Child Age Group."

Q'

=

the release rate of radionuclide i from gaseous release pathway v, in jiCi/s.

For the purpose of implementing the controls of Section 3.1.2.b, only 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation.

All other terms were defined previously.

3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas at or beyond the SITE BOUNDARY due to releases of noble gases from each unit shall be calculated as follows (adapted from Reference 1, page 28, by including only long-term releases):

D8= 3.1 7x10 8 X {(X/IQ)v, Y[NQ~]

(3.13)

Dr =3.17 xl08'y nl(/Qo.XM~i]

(3.14) where:

3.17 x 10- = a units conversion factor: 1 y/(3.15 x 10C s).

Dp

=

the air dose due to beta emissions from noble gas radionuclides, in mrad.

D

=

the air dose due to gamma emissions from noble gas radionuclides, in mrad.

N,

=

the air dose factor due to beta emissions from noble gas radionuclide i (mrad/y)I(giCVm 3), from Table 3-5.

M*

=

the air dose factor due to gamma emissions from noble gas radionuclide i (mrad/y)/(glCi/m 3), from Table 3-5.

N

=

the cumulative release of noble gas radionuclide i from release pathway v

(!iCi), during the period of interest.

and all other terms are as defined above.

Rev. 18 3-31

VEGP ODCM Because the air dose limit is on a per-reactor-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned to the two units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit.

The gaseous final release pathways at the plant site, and the (X/Q),b for each, are identified in Table 3-4.

3-32 Rev. 18 Rev. 18 3-32

VEGP ODOM Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases All values in this table were obtained from Reference 3 (Table B-i), with units converted.

3-33 Rev. 18 Table 3-5.

y - Body (K)

- Skin (L) y,- Air (M) 13-Air (N)

Nuclide (mrem/y) per (mrem/y) per (mrad/y) per (mrad/y) per (pci/m 3)

(ACi/m 3)

(pci/m3)

(90Ci/m

3)

Kr-83m 7.56 E-02 0.00 E+00 1.93 E+01 2.88 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-1 37 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 VEGP ODCM Rev. 18 3-33

VEGP ODOM Table 3-6.

Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated Finite Plume The contents of this table are not applicable to VEGP.

3-34 Rev. 18 VEGP ODCM Rev. 18 3-34

VEGP ODCM 3.4.3 Dose to a Member of the Public at or Beyond Site Boundary The dose received by an individual due to gaseous releases from each reactor unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The MEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the controlling receptor. The dosimetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7.

Doses to a MEMBER OF THE PUBLIC due to gaseous releases of 1-131,1-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases):

Di =.17x08{~aPJ~[v i v]

(3.15) where:

D*

=

the dose to organ j of an individual in age group a, due to gaseous releases of 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in mrem.

3.17 x l0- =a units conversion factor: 1 y/(3.15 x 107 s).

Raipj the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. For the purpose of implementing the controls of Section 3.1.4, the exposure pathways applicable to calculating the dose to the currently-defined controlling receptor are included in Table 3-7; values of Rap for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are included in Tables 3-8 through 3-12.

A detailed discussion of the methods and parameters used for calculating Ra* for the plant site is presented in Chapter 9. That information may be used for recalculating the Rip values if the underlying parameters change, or for calculating Ra values for special radionuclides and age groups when performing the assessments discussed in Section 3.4.4 below.

Wv,

=

the annual average relative dispersion or deposition at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i.

For all tritium pathways, and for the inhalation of any radionuclide: Wip is (X/Q)v, the annual average relative dispersion factor for release pathway v, at the location of the controlling receptor (s/m3). For the ground-plane exposure pathway, and for all ingestion-related pathways for radionuclides other than tritium: Wv is (D/Q),p, the annual average relative deposition factor for release pathway v, at the location of the controlling receptor (m-).

Values of (X/Q)vp and (-DIQX., for use in calculating the dose to the currently defined controlling receptor are included in Table 3-7.

3-35 Rev. 18 Rev. 18 3-35

VEGP ODCM

=

the cumulative release of radionuclide i from release pathway v, during the period of interest (iCi). For the purpose of implementing the controls of Section 3.1.4, only 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation.

In any dose assessment using the methods of this sub-section, only radio nuclides detectable above background in their respective samples should be included in the calculation.

Because the member of the public dose limit is on a per-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one or the other unit.

The gaseous final release pathways at the plant site, and the release elevation for each, are identified in Table 3-4.

3-36 Rev. 18 Rev. 18 3-36

VEGP ODCM ThhIlA *-7.

Attributes of the Controllina Receptor The locations of members of the public in the vicinity of the plant site, and the exposure pathways associated with those locations, are determined in the Annual Land Use Census.

Dispersion and deposition values were calculated based on site meteorological data collected for the period January 1, 1985 through December 31, 1987.

Based on an analysis of this information, the current controlling receptor for the plant site is described as follows (References 15 and 22).

Sector:

Distance:

Aae Group:

SSW 4.7 miles Child Exposure Pathways: Inhalation, ground plane, cow meat, and garden vegetation Dispersion Factors (X/Q)vp:

Ground-Level release points:

Mixed-Mode release points:

7.78 x 108 s/im3 3.78 x 108 s/m3 Deposition Factors (D/Q) P:

Ground-Level release points:

Mixed-Mode release points:

3.43 x 10 m2 1.99 x 101 m2 3-37 Rev. 18 Rev. 18 3-37

VEGP ODCM 3.4.4 Dose Calculations to Support Other Requirements Case 1:

A radiological impact assessment may be required to support evaluation of a reportable event.

Dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters [(X/Q) and (D/Q)] for the period covered by the report, and using the appropriate pathway dose factors (Raipj) for the receptor of interest. Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8.

Values of Raipj other than those presented in Tables 3-8 through 3-12 may need to be calculated. Methods and parameters for calculating values of Raipj are presented in Chapter 9. When calculating Ripj for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the values in Chapter 9, if the specific values are known.

Case 2:

A dose calculation is required to evaluate the results of the Land Use Census, under the provisions of Section 4.1.2.

In the event that the Land Use Census reveals that exposure pathways have changed at previously-identified locations, or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be designated as the controlling receptor. Such dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the annual average dispersion and deposition values

[(X/Q) and (D--7)] for the locations of interest, and using the appropriate pathway dose factors (RaIpj) for the receptors of interest.

Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of Raip other than those presented in Tables 3-8 through 3-12 may need to be calculated. Methods and parameters for calculating values of Raipj are presented in Chapter 9.

Case 3:

Under Section 5.2, a dose calculation may be required to support the determination of a component of the total dose to a receptor other than that currently defined as the controlling receptor.

Dose calculations would be performed using the equations in Section 3.4.3, with the dispersion and deposition parameters and appropriate values of (Rjpj) for the receptor of interest.

Appropriate values of the dispersion and deposition parameters, if not found in Table 3-7, would need to be calculated. Methods for calculating (X/Q) and (D/Q) from meterological data are presented in Chapter 8.

Appropriate values of RaI, if not found in Tables 3-8 through 3-12, would need to be calculated. Methods and parameters for calculating values of Raipj are presented in Chapter 9.

3-38 Rev. 18 Rev. 18 3-38

VEGP ODCM Table 3-8.

Raipo for Ground Plane Pathway, All Age Groups Nuclide T. Body Skin H-3 0.00 0.00 C-14 0.00 0.00 P-32 0.00 0.00 Cr-51 4.66E+06 5.51 E+06 Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 Fe-59 2.73E+08 3.21 E+08 Co-58 3.79E+08 4.44E+08 Co-60 2.15E+10 2.53E+10 Ni-63 0.00 0.00 Zn-65 7.47E+08 8.59E+08 Rb-86 8.99E+06 1.03E+07 Sr-89 2.16E+04 2.51 E+04 Sr-90 0.00 0.00 Y-91 1.07E+06 1.21 E+06 Zr-95 2.45E+08 2.84E+08 Nb-95 1.37E+08 1.61 E+08 Ru-103 1.08E+08 1.26E+08 Ru-106 4.22E+08 5.07E+08 Ag-110in 3.44E+09 4.01 E+09 Sb-124 5.98E+08 6.90E+08 Sb-125 2.34E+09 2.64E+09 Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.08E+05 Te-129m 1.98E+07 2.31 E+07 1-131 1.72E+07 2.09E+07 1-133 2.45E+06 2.98E+06 Cs-134 6.86E+09 8.OOE+09 Cs-136 1.51 E+08 1.71 E+08 Cs-1 37 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Ce-141 1.37E+07 1.54E+07 Ce-1 44 6.95E+07 8.04E+07 Pr-143 0.00 0.00 Nd-147 8.39E+06 1.01 E+07 1.

Units are m2.(mrem/yr)/(p.Ci/s).

2.

The values in the Total Body column also apply to the Bone, Liver, Thyroid, Kidney, Lung, and GI-LLI organs.

3.

This table also supports the calculations of section 6.2.

Rev. 18 3-39

VEGP ODCM Table 3-9.

R.j., for Inhalation Pathway, Child Age Group

[Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 C-14 3.59E+04 6.73E+03 6.73E+03 6.73E+03 6.73E+03 6.73E+03 6.73E+03 P-32 2.60E+06 1.14E+05 9.88E+04 0.00 0.00 0.00 4.22E+04 Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 Mn-54 0.00 4.29E+04 9.51 E+03 0.00 1.00E+04 1.58E+06 2.29E+04 Fe-55 4.74E+04 2.52E+04 7.77E+03 0.00 0.00 1.11 E+05 2.87E+03 Fe-59 2.07E+04 3.34E+04 1.67E+04 0.00 0.00 1.27E+06 7.07E+04 Co-58 0.00 1.77E+03 3.16E+03 0.00 0.00 1.11 E+06 3.44E+04 Co-60 0.00 1.31 E+04 2.26E+04 0.00 0.00 7.07E+06 9.62E+04 Ni-63 8.21 E+05 4.63E+04 2.80E+04 0.00 0.00 2.75E+05 6.33E+03 Zn-65 4.26E+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05 1.63E+04 Rb-86 0.00 1.98E+05 1.14E+05 0.00 0.00 0.00 7.99E+03 Sr-89 5.99E+05 0.00 1.72E+04 0.00 0.00 2.16E+06 1.67E+05 Sr-90 1.01 E+08 0.00 6.44E+06 0.00 0.00 1.48E+07 3.43E+05 Y-91 9.14E+05 0.00 2.44E+04 0.00 0.00 2.63E+06 1.84E+05 Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.96E+04 2.23E+06 6.11 E+04 Nb-95 2.35E+04 9.18E+03 6.55E+03 0.00 8.62E+03 6.14E+05 3.70E+04 Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 6.62E+05 4.48E+04 Ru-106 1.36E+05 0.00 1.69E+04 0.00 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+06 1.OOE+05 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 6.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 1-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00 2.84E+03 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.00 5.48E+03 Cs-134 6.51 E+05 1.01 E+06 2.25E+05 0.00 3.30E+05 1.21 E+05 3.85E+03 Cs-136 6.51E+04 1.71E+05 1.16E+05 0.00 9.55E+04 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11 E+01 1.74E+06 1.02E+05 Ce-141 3.92E+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.66E+04 Ce-144 6.77E+06 2.12E+06 3.611E+05 0.00 1.17E+06 1.20E+07 3.89E+05 Pr-143 1.85E+04 5.55E+03 9.14E+02 0.00 3.OOE+03 4.33E+05 9.73E+04 Nd-147 1.08E+04 8.73E+03 6.81 E+02 0.00 4.81 E+03 3.28E+05 8.21 E+04 Units are (mrem/yr)/(pxCi/m 3) for all radionuclides.

This table also supports the calculations of section 6.2.

3-40 Rev. 18

1.

2.

Rev. 18 3-40

VEGP ODCM Table 3-10.

Ra*ip for Inhalation Pathway, Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 C-14 1.82E+04 3.41 E+03 3.41 E+03 3.41 E+03 3.41 E+03 3.41 E+03 3.41 E+03 P-32 1.32E+06 7.71 E+04 5.01 E+04 0.00 0.00 0.00 8.64E+04 Cr-51 0.00 0.00 1.00E+02 5.95E+01 2.28E+01 1.44E+04 3.32E+03 Mn-54 0.00 3.96E+04 6.30E+03 0.00 9.84E+03 1.40E+06 7.74E+04 Fe-55 2.46E+04 1.70E+04 3.94E+03 0.00 0.00 7.21 E+04 6.03E+03 Fe-59 1.18E+04 2.78E+04 1.06E+04 0.00 0.00 1.02E+06 1.88E+05 Co-58 0.00 1.58E+03 2.07E+03 0.00 0.00 9.28E+05 1.06E+05 Co-60 0.00 1.15E+04 1.48E+04 0.00 0.00 5.97E+06 2.85E+05 Ni-63 4.32E+05 3.14E+04 1.45E+04 0.00 0.00 1.78E+05 1.34E+04 Zn-65 3.24E+04 1.03E+05 4.66E+04 0.00 6.90E+04 8.64E+05 5.34E+04 Rb-86 0.00 1.35E+05 5.90E+04 0.00 0.00 0.00 1.66E+04 Sr-89 3.04E+05 0.00 8.72E+03 0.00 0.00 1.40E+06 3.50E+05 Sr-90 9.92E+07 0.00 6.10E+06 0.00 0.00 9.60E+06 7.22E+05 Y-91 4.62E+05 0.00 1.24E+04 0.00 0.00 1.70E+06 3.85E+05 Zr-95 1.07E+05 3.44E+04 2.33E+04 0.00 5.42E+04 1.77E+06 1.50E+05 Nb-95 1.41 E+04 7.82E+03 4.21 E+03 0.00 7.74E+03 5.05E+05 1.04E+05 Ru-103 1.53E+03 0.00 6.58E+02 0.00 5.83E+03 5.05E+05 1.10E+05 Ru-106 6.91 E+04 0.00 8.72E+03 0.00 1.34E+05 9.36E+06 9.12E+05 Ag-110m 1.08E+04 1.OOE+04 5.94E+03 0.00 1.97E+04 4.63E+06 3.02E+05 Sb-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 0.00 2.48E+06 4.06E+05 Sb-125 6.61E+04 7.13E+02 1.33E+04 5.87E+01 0.00 2.20E+06 1.01E+05 Te-125m 3.42E+03 1.58E+03 4.67E+02 1.05E+03 1.24E+04 3.14E+05 7.06E+04 Te-1 27m 1.26E+04 5.77E+03 1.57E+03 3.29E+03 4.58E+04 9.60E+05 1.50E+05 Te-129m 9.76E+03 4.67E+03 1.58E+03 3.44E+03 3.66E+04 1.16E+06 3.83E+05 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 0.00 6.28E+03 1-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 0.00 8.88E+03 Cs-134 3.73E+05 8.48E+05 7.28E+05 0.00 2.87E+05 9.76E+04 1.04E+04 Cs-136 3.90E+04 1.46E+05 1.10E+05 0.00 8.56E+04 1.20E+04 1.1 7E+04 Cs-1 37 4.78E+05 6.21 E+05 4.28E+05 0.00 2.22E+05 7.52E+04 8.40E+03 Ba-140 3.90E+04 4.90E+01 2.57E+03 0.00 1.67E+01 1.27E+06 2.18E+05 Ce-141 1.99E+04 1.35E+04 1.53E+03 0.00 6.26E+03 3.62E+05 1.20E+05 Ce-144 3.43E+06 1.43E+06 1.84E+05 0.00 8.48E+05 7.78E+06 8.16E+05 Pr-143 9.36E+03 3.75E+03 4.64E+02 0.00 2.16E+03 2.81 E+05 2.00E+05 Nd-147 5.27E+03 6.1OE+03 3.65E+02 0.00 3.56E+03 2.21 E+05 1.73E+05 Units are (mrem/yr)/(pgCi/m 3) for all radionuclides.

This table is included to support the calculations of section 6.2.

3-41 Rev. 18

1.

2.

3-41 Rev. 18

VEGP ODCM Table 3-11.

Ri,, for Cow Meat Pathway, Child Age Group Nuclide Bone Liver T. Body I Thyroid Kidney Lung GI-LLI H-3 0.00 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 C-14 5.29E+05 1.06E+05 1.06E+05 1.06E+05 1.06E+05 1.06E+05 1.06E+05 P-32 7.41 E+09 3.47E+08 2.86E+08 0.00 0.00 0.00 2.05E+08 Cr-51 0.00 0.00 8.79E+03 4.88E+03 1.33E+03 8.91 E+03 4.66E+05 Mn-54 0.00 8.01E+06 2.13E+06 0.00 2.25E+06 0.00 6.72E+06 Fe-55 4.57E+08 2.42E+08 7.51 E+07 0.00 0.00 1.37E+08 4.49E+07 Fe-59 3.76E+08 6.09E+08 3.03E+08 0.00 0.00 1.77E+08 6.34E+08 Co-58 0.00 1.64E+07 5.02E+07 0.00 0.00 0.00 9.58E+07 Co-60 0.00 6.93E+07 2.04E+08 0.00 0.00 0.00 3.84E+08 Ni-63 2.91E+10 1.56E+09 9.91E+08 0.00 0.00 0.00 1.05E+08 Zn-65 3.75E+08 1.OOE+09 6.22E+08 0.00 6.30E+08 0.00 1.76E+08 Rb-86 0.00 5.77E+08 3.55E+08 0.00 0.00 0.00 3.71 E+07 Sr-89 4.82E+08 0.00 1.38E+07 0.00 0.00 0.00 1.87E+07 Sr-90 1.04E+1 0 0.00 2.64E+09 0.00 0.00 0.00 1.40E+08 Y-91 1.80E+06 0.00 4.82E+04 0.00 0.00 0.00 2.40E+08 Zr-95 2.66E+06 5.85E+05 5.21 E+05 0.00 8.38E+05 0.00 6.11 E+08 Nb-95 3.1OE+06 1.21E+06 8.62E+05 0.00 1.13E+06 0.00 2.23E+09 Ru-103 1.55E+08 0.00 5.96E+07 0.00 3.90E+08 0.00 4.01 E+09 Ru-106 4.44E+09 0.00 5.54E+08 0.00 5.99E+09 0.00 6.90E+10 Ag-110in 8.39E+06 5.67E+06 4.53E+06 0.00 1.06E+07 0.00 6.74E+08 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 5.69E+08 1.54E+08 7.59E+07 1.60E+08 0.00 0.00 5.49E+08 Te-127m 1.77E+09 4.78E+08 2.11 E+08 4.24E+08 5.06E+09 0.00 1.44E+09 Te-129m 1.79E+09 5.OOE+08 2.78E+08 5.77E+08 5.26E+09 0.00 2.18E+09 1-131 1.65E+07 1.66E+07 9.46E+06 5.50E+09 2.73E+07 0.00 1.48E+06 1-133 5.67E-01 7.02E-01 2.66E-01 1.30E+02 1.17E+00 0.00 2.83E-01 Cs-134 9.22E+08 1.51 E+09 3.19E+08 0.00 4.69E+08 1.68E+08 8.16E+06 Cs-136 1.62E+07 4.46E+07 2.88E+07 0.00 2.37E+07 3.54E+06 1.57E+06 Cs-137 1.33E+09 1.28E+09 1.88E+08 0.00 4.16E+08 1.50E+08 7.99E+06 Ba-140 4.38E+07 3.84E+04 2.56E+06 0.00 1.25E+04 2.29E+04 2.22E+07 Ce-141 2.22E+04 1.11E+04 1.64E+03 0.00 4.86E+03 0.00 1.38E+07 Ce-144 2.32E+06 7.26E+05 1.24E+05 0.00 4.02E+05 0.00 1.89E+08 Pr-1 43 3.34E+04 1.OOE+04 1.66E+03 0.00 5.43E+03 0.00 3.60E+07ý Nd-147 1.17E+04 9.47E+03 7.33E+02 0.00 5.19E+03 0.00 1.50E+07 Units are (mrem/yr)/(glCi/m 3) for tritium, and m2.(mrem/yr)/(l.Ci/S) for all other radionuclides.

Rev. 18 3-42

VEGP ODCM Table 3-12.

Ro0 o, for Garden Veaetation Pathway, Child Age Group Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI H-3 0.00 4.01 E+03 4.01 E+03 4.01 E+03 4.01 E+03 4.01 E+03 4.01 E+03 C-14 8.89E+08 1.78E+08 1.78E+08 1.78E+08 1.78E+08 1.78E+08 1.78E+08 P-32 3.37E+09 1.58E+08 1.30E+08 0.00 0.00 0.00 9.31 E+07 Cr-51 0.00 0.00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+06 Mn-54 0.00 6.65E+08 1.77E+08 0.00 1.86E+08 0.00 5.58E+08 Fe-55 8.01 E+08 4.25E+08 1.32E+08 0.00 0.00 2.40E+08 7.87E+07 Fe-59 3.98E+08 6.43E+08 3.20E+08 0.00 0.00 1.86E+08 6.70E+08 Co-58 0.00 6.44E+07 1.97E+08 0.00 0.00 0.00 3.76E+08 Co-60 0.00 3.78E+08 1.12E+09 0.00 0.00 0.00 2.1OE+09 Ni-63 3.95E+10 2.11E+09 1.34E+09 0.00 0.00 0.00 1.42E+08 Zn-65 8.13E+08 2.16E+09 1.35E+09 0.00 1.36E+09 0.00 3.80E+08 Rb-86 0.00 4.52E+08 2.78E+08 0.00 0.00 0.00 2.91 E+07 Sr-89 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-90 1.24E+12 0.00 3.15E+11 0.00 0.00 0.00 1.67E+10 Y-91 1.86E+07 0.00 4.99E+05 0.00 0.00 0.00 2.48E+09 Zr-95 3.86E+06 8.48E+05 7.55E+05 0.00 1.21 E+06 0.00 8.85E+08 Nb-95 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.96E+08 Ru-103 1.53E+07 0.00 5.90E+06 0.00 3.86E+07 0.00 3.97E+08 Ru-106 7.45E+08 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ag-110m 3.21 E+07 2.17E+07 1.73E+07 0.00 4.04E+07 0.00 2.58E+09 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 3.51 E+08 9.50E+07 4.67E+07 9.84E+07 0.00 0.00 3.38E+08 Te-127m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00 1.07E+09 Te-1 29m 8.41 E+08 2.35E+08 1.31 E+08 2.71 E+08 2.47E+09 0.00 1.03E+09 1-131 1.43E+08 1.44E+08 8.17E+07 4.75E+10 2.36E+08 0.00 1.28E+07 1-133 3.53E+06 4.37E+06 1.65E+06 8.11 E+08 7.28E+06 0.00 1.76E+06 Cs-134 1.60E+10 2.63E+10 5.55E+09 0.00 8.15E+09 2.93E+09 1.42E+08 Cs-1 36 8.24E+07 2.27E+08 1.47E+08 0.00 1.21 E+08 1.80E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.38E+09 0.00 7.46E+09 2.68E+09 1.43E+08 Ba-140 2.77E+08 2.42E+05 1.61 E+07 0.00 7.89E+04 1.45E+05 1.40E+08 Ce-141 6.56E+05 3.27E+05 4.86E+04 0.00 1.43E+05 0.00 4.08E+08 Ce-144 1.27E+08 3.98E+07 6.78E+06 0.00 2.21 E+07 0.00 1.04E+10 Pr-143 1.46E+05 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.57E+08 Nd-147 7.15E+04 5.79E+04 4.48E+03 0.00 3.18E+04 0.00 9.17E+07 Units are (mrem/yr)/(RCi/m 3) for tritium, and m2-(mrem/yr)/(OCi/s) for all other radionuclides.

Rev. 18 3-43

VEGP ODCM 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3.5.1 Thirty-One Day Dose Projections In order to meet the requirements of the limit for operation of the gaseous radwaste treatment system (see Section 3.1.5), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to areas at or beyond the SITE BOUNDARY of gaseous effluents containing radioactive materials occurs or is expected.

Projected 31-day air doses and doses to individuals due to gaseous effluents may be determined as follows:

For air doses:

(3.16)

For individual doses:

Dop =(Doc 1x31+D (3.17) where:

Dpp

=

the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases.

DR

=

the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

Do

=

the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Do, may be set to zero.

D'

=

the projected air dose due to gamma emissions from noble gases for the next 31 days of gaseous releases.

DI

=

the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

D*

=

the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine Rev. 18 3-44

VEGP ODCM gaseous effluents. If only routine gaseous effluents are anticipated, D. may be set to zero.

Dop

=

the projected dose to the total body or organ o, due to releases of 1-131, 1-133, tritium, and particulates for the next 31 days of gaseous releases.

Doc

=

the cumulative dose to the total body or organ o, due to releases of 1-131, 1-133, tritium, and particulates that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

Doa

=

the anticipated dose to the total body or organ o, due to releases of 1-131, 1-133, tritium, and particulates, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Doa may be set to zero.

t

=

the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).

3.5.2 Dose Proiections for Specific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For air dose and individual dose projections due to gaseous effluent releases, follow the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to exist during the release period.

3-45 Rev. 18 Rev. 18 3-45

VEGP ODCM 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS Section of Term Definition Initial Use AG =

the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit among all the release pathways [unitless].

3.3.2.1 AG, =

the administrative allocation factor for gaseous source stream s, applied to divide the gaseous release limit among all the release pathways [unitless].

3.3.3 AGv =

the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous release limit among all the release pathways [unitless].

3.3.2.2 c =

the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prior to release

[9CVmL].

3.3.2.1 Cns =

the calculated noble gas effluent monitor setpoint for gaseous source stream s [jiCi/mL].

3.3.3 cnv =

the calculated noble gas effluent monitor setpoint for release pathway v [CiCVmL].

3.3.2.2 Dja =

the dose to organ j of an individual in age group a, due to gaseous releases of 1-131,1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days [mrem].

3.4.3 Doa =

the anticipated dose to organ o due to releases of non-noble-gas radionuclides, contributed by any planned activities during the next 31-day period [mrem].

3.5.1 Do, =

the cumulative dose to organ o due to releases of non-noble-gas radionuclides that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrem].

3.5.1 Dop =

the projected dose to organ o due to the next 31 days of gaseous releases of non-noble-gas radionuclides [mrem].

3.5.1 Dp =

the air dose due to beta emissions from noble gas radionuclides

[mrad].

3.4.2 D=

the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period [mrad].

3.5.1 3-46 Rev. 18 Rev. 18 3-46

VEGP ODCM Term DI* =

3-47 Rev. 18 Definition the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad].

the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases [mrad].

the air dose due to gamma emissions from noble gas radionuclides [mrad].

the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period [mrad].

the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad].

the projected air dose due to gamma emissions from noble gases, for the next 31 days of gaseous releases [mrad].

the annual average relative deposition factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [m 2].

the skin dose rate at the time of the release [mrem/y].

the dose rate to organ o at the time of the release [mrem/y].

the total body dose rate at the time of the release [mrem/y].

the maximum anticipated actual discharge flowrate for release pathway v during the period of the planned release [mL/s].

the maximum anticipated actual discharge flowrate for gaseous source stream s during the period of the planned release [mL's].

the total body dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrem/y)/(R.Ci/m 3)].

the skin dose factor due to beta emissions from noble gas radio nuclide i, from Table 3-5 [(mrem/y)/(gCi/m3)].

the air dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrad/y)/(gCi/m3)].

Dy

=

(D/Q)v~

DRk =

DRo=

DRt =

fav =

fas =

Section of Initial Use 3.5.1 3.5.1 3.4.2 3.5.1 3.5.1 3.5.1 3.4.3 3.4.1.1 3.4.1.2 3.4.1.1 3.3.2.2 3.3.3 3.3.2.2 3.3.2.2 3.4.2 Rev. 18 3-47

VEGP ODCM Section of Term Definition Initial Use N =

the number of simultaneously active gaseous release pathways

[unitless].

3.3.4 N, =

the air dose factor due to beta emissions from noble gas radio nuclide i, from Table 3-5 [(mrad/y)/(p.Ci/m3)].

3.4.2 Pi. =

the site-specific dose factor for radionuclide i (1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) and organ o. The values of P* are equal to the site specific Raipj values presented in Table 3-9 [(mrem/y)/(! Ci/m 3)].

3.4.1.2 oil =

the release rate of noble gas radionuclide i from release pathway v during the period of interest [gCi/s].

3.3.2.2

a. i,=

the release rate of radionuclide i (1-131,1-133, tritium, and radio nuclides in particulate form with half-lives greater than 8 days) from gaseous release pathway v during the period of interest

[!9Ci/s].

3.4.1.2 4=

the cumulative release of noble gas radionuclide i from release pathway v during the period of interest [pCi].

3.4.2 OI,=

the cumulative release of non-noble-gas radionuclide i from release pathway v, during the period of interest [pCi].

3.4.3 Raopi =

the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. Values and units of Rjpj for each exposure pathway, age group, and radionuclide that may arise in calculations for implementing Section 3.1.4 are listed in Table 3-8 through Table 3-9.

3.4.3 Rk =

the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the release under consideration

[unitless].

3.3.2.1 Rt =

the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the release under consideration [unitless].

3.3.2.1 rk =

the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the source stream under consideration [unitless].

3.3.3.1 rt =

the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration [unitless].

3.3.3.1 Rev. 18 3-48

Term SF =

t =

Wvip =

X =

Xir =

Xis I XIS =

(X/Q) =

(X/Q) b =

(X/Q),b =

(X/Q) =

Definition the safety factor used in gaseous setpoint calculations to compensate for statistical fluctuations and errors of measurement

[unitless].

the number of whole or partial days elapsed in the current quarter, including the period of the release under consideration.

the annual average relative dispersion [(X/Q)L] or deposition

[(D/Q)vp] at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i.

the noble gas concentration for the release under consideration

[CiCVmL].

the concentration of radionuclide i applicable to active gaseous release pathway r [tCi/mL].

the measured concentration of radionuclide i in gaseous source stream s [gCVmL].

the measured concentration of radionuclide i in gaseous stream v

[ILCi/mL].

the highest relative concentration at any point at or beyond the SITE BOUNDARY [s/m3].

the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r [s/m 3].

the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v, from Table 3-4 [s/m3].

annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [s/m3].

3-49 Rev. 18 VEGP ODCM Section of Initial Use 3.3.2.2 3.5.1 3.4.3 3.3.2.1 3.3.4 3.3.3 3.3.2.2 3.3.2.1 3.3.4 3.3.2.2 3.4.3 Rev. 18 3-49