NL-13-1006, Annual Radiological Environmental Operating Reports for 2012: Difference between revisions
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In addition, several other groups within Southern Company are now utilized to conduct an improved sampling program and to provide additional expertise in characterizing groundwater quality and flow. The sampling frequency for radiological groundwater monitoring was officially changed to quarterly starting in second quarter of 2008 with SCS Civil Field Services perfonning the sampling and Georgia Power Environmental Laboratory continuing to analyze the samples. | In addition, several other groups within Southern Company are now utilized to conduct an improved sampling program and to provide additional expertise in characterizing groundwater quality and flow. The sampling frequency for radiological groundwater monitoring was officially changed to quarterly starting in second quarter of 2008 with SCS Civil Field Services perfonning the sampling and Georgia Power Environmental Laboratory continuing to analyze the samples. | ||
Southern Nuclear Corporate Engineering and Hatch Site Engineering have developed a Buried Piping and Tanks Inspection Program. This program should help to prevent releases of radioactive material to groundwater. Underground piping and components are risked ranked using detailed procedures and EPRI's software, BPWorks, to ensure vulnerable areas are identified and repaired or replaced before problems occur. | Southern Nuclear Corporate Engineering and Hatch Site Engineering have developed a Buried Piping and Tanks Inspection Program. This program should help to prevent releases of radioactive material to groundwater. Underground piping and components are risked ranked using detailed procedures and EPRI's software, BPWorks, to ensure vulnerable areas are identified and repaired or replaced before problems occur. | ||
In May of 2009, there was an increase in tritium concentration in well T-3 (located near the U-l Turbine Building) from approximately 2600 pCi/1 to approximately 37,000 pCi/1. Neighboring well N9B (not part of the formal GW sampling program) also showed an approximate lOX increase - going from 1300 pCi/1 to over 10K pCi/l. Investigation found no process leaks and the non-rad constituents continued to match groundwater. The increase was attributed to migration of the plume. Increased rainfall and the fact that the wells are located near the subsurface drain could likely have facilitated the pathway of the plume towards the T3 well. A courtesy notification was made to the State of Georgia Dept. of Natural Resources, and a | In May of 2009, there was an increase in tritium concentration in well T-3 (located near the U-l Turbine Building) from approximately 2600 pCi/1 to approximately 37,000 pCi/1. Neighboring well N9B (not part of the formal GW sampling program) also showed an approximate lOX increase - going from 1300 pCi/1 to over 10K pCi/l. Investigation found no process leaks and the non-rad constituents continued to match groundwater. The increase was attributed to migration of the plume. Increased rainfall and the fact that the wells are located near the subsurface drain could likely have facilitated the pathway of the plume towards the T3 well. A courtesy notification was made to the State of Georgia Dept. of Natural Resources, and a 10CFR50.72 formal report was made to the NRC - although only courtesy notifications were required per procedure. | ||
In 2012, the quarterly tritium concentrations in T3 ranged from 1840 pCi/1 to 6320 pC il l during the year (average of 3282 pCi/1 in 2012) but remained below the established Administative Control Limit (ACL) of 37,000 pCi/1. Administrative Control Limits (ACL) were established near the end of 20 I 0 for the surficial and deep aquifers and for specific wells based on the presence of legacy tritium, the previous well results, and total measurement uncertainty. There are no reporting requirements associated with exceeding an ACL but additional actions would be taken to verify no new sources of tritium if an ACL was exceeded. The ACL for T-12 is 900,000 pCi/1 and the average for T-12 in 2012 was 99,000 pCi/1 of tritium with a range in values of 50,500 to 212,000 pCi/1 (157,775 pCi/ 1 was the 4-43 | In 2012, the quarterly tritium concentrations in T3 ranged from 1840 pCi/1 to 6320 pC il l during the year (average of 3282 pCi/1 in 2012) but remained below the established Administative Control Limit (ACL) of 37,000 pCi/1. Administrative Control Limits (ACL) were established near the end of 20 I 0 for the surficial and deep aquifers and for specific wells based on the presence of legacy tritium, the previous well results, and total measurement uncertainty. There are no reporting requirements associated with exceeding an ACL but additional actions would be taken to verify no new sources of tritium if an ACL was exceeded. The ACL for T-12 is 900,000 pCi/1 and the average for T-12 in 2012 was 99,000 pCi/1 of tritium with a range in values of 50,500 to 212,000 pCi/1 (157,775 pCi/ 1 was the 4-43 | ||
2011 average in T-12). For NW-I0, the ACL is 160,000 pCi/\ and the average tritium concentration in 2012 was 15,075 pCi/\ with a range of 12,000 to 17,600 pCi/1 (22,225 pCi/1 was the average in 20 II). These two wells, in areas of legacy contamination, continue to trend downward. | 2011 average in T-12). For NW-I0, the ACL is 160,000 pCi/\ and the average tritium concentration in 2012 was 15,075 pCi/\ with a range of 12,000 to 17,600 pCi/1 (22,225 pCi/1 was the average in 20 II). These two wells, in areas of legacy contamination, continue to trend downward. | ||
The ACL for T 10 (25,000 pCi/l) was exceeded on 9/21 Ill. The result, which was detennined on 9/28111, was 4.61E6 pCi/1 and the previous sample had been only 5000 pCi/1. Additional samples were taken to verify the high tritium level and hard to detect radionuclides, Sr-89/90 and Fe-55, were analyzed for but were detennined to be at background levels. No gamma emitters were detected. | The ACL for T 10 (25,000 pCi/l) was exceeded on 9/21 Ill. The result, which was detennined on 9/28111, was 4.61E6 pCi/1 and the previous sample had been only 5000 pCi/1. Additional samples were taken to verify the high tritium level and hard to detect radionuclides, Sr-89/90 and Fe-55, were analyzed for but were detennined to be at background levels. No gamma emitters were detected. | ||
Voluntary communications with state/local stakeholders were perfonned on 9/29111. A fonnal | Voluntary communications with state/local stakeholders were perfonned on 9/29111. A fonnal 10CFR50. 72b notification to NRC was also made. A response team was assembled (with 24 hour coverage) to identify the source of the tritium. | ||
Buried piping that supplied water from the CST -I to the liquid radwaste processing system was identified as the source, and the section of piping was evacuated of water and abandoned in place. A design change was completed to replace it with aboveground piping. No drinking water sources were impacted by the leak, and the tritium was contained within the shallow perched aquifer located in the immediate vicinity of CST-I. Two new wells, NU-1 and NU-2, were drilled in the area near the leak to enhance monitoring and to facilitate remediation activities (pumping out wells to remove the contamination). By 4th quarter sampling (12/06111), the tritium in T-1O had decreased to 2.3E6 pCi/1. | Buried piping that supplied water from the CST -I to the liquid radwaste processing system was identified as the source, and the section of piping was evacuated of water and abandoned in place. A design change was completed to replace it with aboveground piping. No drinking water sources were impacted by the leak, and the tritium was contained within the shallow perched aquifer located in the immediate vicinity of CST-I. Two new wells, NU-1 and NU-2, were drilled in the area near the leak to enhance monitoring and to facilitate remediation activities (pumping out wells to remove the contamination). By 4th quarter sampling (12/06111), the tritium in T-1O had decreased to 2.3E6 pCi/1. | ||
Sample results for this location in 2012 have averaged 61,766 pCi/1. | Sample results for this location in 2012 have averaged 61,766 pCi/1. | ||
Revision as of 14:52, 11 November 2019
| ML13136A048 | |
| Person / Time | |
|---|---|
| Site: | Hatch, Vogtle, Farley  |
| Issue date: | 05/15/2013 |
| From: | Marino P Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-13-1006 | |
| Download: ML13136A048 (267) | |
Text
{{#Wiki_filter:Paula M. Marino Southern Nuclear Vice President Operating Company. Inc Engineering 40 Inverness Center Parkway Birmingham, Alabama 35242 Tel 205.992.7707 Fax 205.992.6165 pmmarino@southernco.com May 15, 2013 SOUTHERN'\ COMPANY Docket Nos.: 50-321 50-348 50-424 NL-13-1006 50-366 50-364 50-425 U. S. Nuclear Regulatory Commission ATfl'J: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Joseph M. Farley Nuclear Plant Vogtle Electric Generating Plant Annual Radiological Environmental Operating Reports for 2012 Ladies and Gentlemen: In accordance with section 5.6.2 of the referenced plants' Technical Specifications, Southern Nuclear Operating Company hereby submits the Annual Radiological Environmental Operating Reports for 2012. This letter contains no NRC commitments. If you have any questions, please advise. Respectfully submitted, Paula M. Marino Vice President Engineering PMM/gal/lac
Enclosures:
- 1. Hatch Annual Radiological Environmental Operating Report for 2012
- 2. Farley Annual Radiological Environmental Operating Report for 2012
- 3. Vogtle Annual Radiological Environmental Operating Report for 2012
U. S. Nuclear Regulatory Commission NL-13-1006 Page 2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. D. R. Madison, Vice President - Hatch Mr. T. E. Tynan, Vice President - Vogtle Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. C. R. Pierce, Regulatory Affairs Director RType : CFA04.054; CHA02 .004; CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch, Vogtle Mr. E. D. Morris, Senior Resident Inspector - Hatch Mr. L. M. Cain, Senior Resident Inspector - Vogtle Ms. E. A. Brown, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Mr. J. R. Sowa, Senior Resident Inspector - Farley State of Alabama Mr. J. L. McNees, Department of Public Health, Office of Radiation Control State of Georgia Mr. M. Williams, Department of Natural Resources American Nuclear Insurers Mr. R. A. Oliveira
Edwin I. Hatch Nuclear Plant Joseph M. Farley Nuclear Plant Vogtle Electric Generating Plant Annual Radiological Environmental Operating Reports for 2012 Enclosure 1 Hatch Annual Radiological Environmental Operating Reports for 2012
EDWIN I. HATCH NUCLEAR PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 SOUTHERN COMPANY A Energy to Serve Your World'"
EDWIN I. HATCH NUCLEAR PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 April 26,2 012 FINAL Chern Tim Meents Ian Lake Tim .M ee nts@ chem st aff.co m lan .Lake@chemstaff.co m 815-600-9247 815-600-2067 Denn is Oltman s DOltm ans @c hemst aff.co m 717-575-3481
T ABLE OF CONTENTS Title and/or Section Subsection Page List of Figures 11
...
List of Tables 111 List of Acronyms IV 1.0 Introduction I-I 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 4-5 4 .2 Airborne 4-7 4.3 Direct Radiation 4-12 4.4 Milk 4-18 4.5 Vegetation 4-22 4 .6 River Water 4-25 4 .7 Fish 4-28 4.8 Sediment 4-33 4.9 Groundwater 4-41 5.0 Interlaboratory Comparison Program 5-1 6.0 Conclusions 6-1 7.0 EITata 7-1
LIST OF FIGURES Figure Number Title Page Figure 2-1 REMP Stations Near the Plant 2-8 Figure 2-2 REMP Stations Beyond Six Miles from the Plant 2-9 Figure 2-3 Groundwater Monitoring Locations 2-10 Figure 2-4 Deep Wells 2-11 Figure 4.2-1 A verage Weekly Gross Beta Air Concentration 4-7 Figure 4.2-2 A verage Annual Cs-137 Concentration in Air 4-9 Figure 4.3-1 A verage Quarterly Exposure from Direct Radiation 4-13 Figure 4.3-2 A verage Quarterly Exposure from Direct Radiation at Special Interest Areas 4-15 Figure 4.4-1 Average Annual Cs-137 Concentration in Milk 4-18 Figure 4.4-2 A verage Annual 1-131 Concentration in Milk 4-20 Figure 4.5-1 Average Annual Cs-137 Concentration in Vegetation 4-23 Figure 4.6-1 A verage Annual H -3 Concentration in River Water 4-26 Figure 4.7-1 Average Annual Cs-137 Concentration in Fish 4-29 Figure 4.7-2 Average Annual Cs-134 Concentration in Fish 4-31 Figure 4.8-1 A verage Annual Co-60 Concentration in Sediment 4-34 Figure 4.8-2 A verage Annual Cs- 137 Concentration in Sediment 4-37 Figure 4.8-3 A verage Annual Indicator Station Concentrations of Select Nuclides in Sediment 4-39 Figure 4.9-1 Plant Hatch Unconfined Perched Aquifer 4th Qtr 2011 4-45 11
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 2-5 Table 2-3 Groundwater Monitoring 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 4-5 Table 4.2-1 Average Weekly Gross Beta Air Concentration 4-8 Table 4.2-2 Average Annual Cs-137 Concentration in Air 4-10 Table 4.3-1 A verage Quarterly Exposure from Direct Radiation 4-14 Table 4.3-2 A verage Quarterly Exposure from Direct Radiation at Special Interest Areas 4-16 Table 4.4-1 A verage Annual Cs-13 7 Concentration in Milk 4-19 Table 4.4-2 Average Annual 1-131 Concentration in Milk 4-21 Table 4.5-1 A verage Annual Cs-13 7 Concentration in Vegetation 4-24 Table 4.6-1 A verage Annual H-3 Concentration in River Water 4-27 Table 4.7-1 Average Annual Cs-137 Concentration in Fish 4-30 Table 4.7-2 Average Annual Cs-134 Concentration in Fish 4-32 Table 4.8-1 A verage Annual Co-60 Concentration in Sediment 4-35 Table 4.8-2 A verage Annual Cs-13 7 Concentration in Sediment 4-38 Table 4 .8-3 Sediment Nuclide Concentrations Other Than Co-60 & Cs-137 4-40 Table 5-1 Interlaboratory Comparison Results 5-3 111
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 HNP Edwin 1. Hatch Nuclear Plant ICP Interlaboratory Comparison Program MDC Minimum Detectable Concentration MDD Minimum Detectable Difference NA Not Applicable NDM No Detectable Measurement(s) NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual OSL Optically Stimulated Luminescence Po Preoperation REMP Radiological Environmental Monitoring Program RL Reporting Level TLD Thermoluminescent Dosimeter TS Technical Specification IV
1.0 INTRODUCTION
The Radiological Environmental Monitoring Program (REMP) is conducted in accordance with Chapter 4 of the Offsite Dose Calculation Manual (ODCM). REMP activities for 2012 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 Edwin 1. Hatch Nuclear Plant (HNP).
The assessments include comparisons between the results of analyses of samples obtained at locations where radiological levels are not expected to be affected by plant operation (control stations), areas of higher population (community 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. The pre-operational stage of the REMP began with the establishment and activation of the environmental monitoring stations in January of 1972. The operational stage of the REMP began on September 12, 1974 with Unit 1 initial criticality. A description of the REMP is provided in Section 2 of this report. An annual summary of the results of the analyses of REMP samples is provided in Section 3. A discussion of the results, including assessments of any radiological impacts upon the environment, and the results of the land use census and the river survey, are provided 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. The sampling locations (stations) specified by ODCM 4.2 are depicted on maps in Figures 2-1 and 2-2. These maps are keyed to Table 2-2 which delineates the direction and distance of each station from the main stack. 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 3)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Approximate Sampling and Type of Analysis and Frequency and/or Sample Number of Sample Collection Frequency Locations I. Airborne 6 Continuous operation Radioiodine canister: I-131 analysis, weekly. Radioiodine and of the sampler with Particulates sample collection Particulate sampler: analyze for gross beta radioactivity not less weekly. than 24 hours following filter change, weekly; perform gamma isotopic analysis on affected sample when gross beta activity is 10 times the yearly mean of control samples; and composite (by location) for gamma isotopic analysis, quarterly.
- 2. Direct Radiation 37 Quarterly Gamma dose, quarterly.
}_.__~g~~tion______ _ _ __._____ _ _ ______ ______ __ ____ .. . __. _ ___ ._. . ._. . _________. . . . . __. ._..._._____.__.__. _. _.__..._... _.._____. ____. . __.. . __ . _._ .__._._._._. __. _. __.__ _.. __. __. . .__._. __ . ..
N I Milk (a) Bimonthly Gamma isotopic and 1-131 analysis, bimonthly. N Fish or Clams (b) 2 Semiannually Gamma isotopic analysis on edible portions, semiannually.
.... __.__.....__ ..... _... _._._._ _... - - _.
Grass or Leafy 3 Monthly during Gamma isotopic analysis, monthly. (c) Vegetation growIng season.
- 4. Waterborne Surface 2 Composite sample Gamma isotopic analysis, monthly. Composite (by location) collected monthly. (d) for tritium analysis, quarterly.
Sediment 2 Semiannually. Gamma isotopic analysis, semiannually.
TABLE 2-1 (SHEET 2 of 3)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Approximate Sampling and Type of Analysis and Frequency and/or Sample Number of Sample Collection Frequency Locations Drinking Water One sample of river River water collected 1-131 analysis on each sample when biweekly collections are (e & f) water near the intake near the intake will be required. Gross beta and gamma isotopic analysis on each and one sample of a composite sample; sample; composite (by location) for tritium analysis, quarterly. finished water from the finished water will each of one to three be a grab sample. of the nearest water These samples will be supplies which could collected monthly be affected by HNP unless the calculated discharges. dose due to consumption of the water is greater than 1
,
tv w mrem/year; then the collection will be biweekly. The collections may revert to monthly should the calculated doses become less than I mrernlyear.
--G~o~-;d;~t~-~-- ---- - - --SeeT~b le-2-=-f-------- -Quartefly-sample;-------Tntlum, ia-mma -lsoto-P1C~ancnleTCrpa-rameters-(pH, --***-*-*-. -.. .----.-.. .
Figure 2-3, and pump used to sample temperature, conductivity, dissolved oxygen, Figure 2-4 GW wells; grab oxidatiolvreduction potential, and turbidity) of each sample sample from yard quarterly; Hard to detect radionuclides as necessary based on drains and ponds results of tritium and gamma
TABLE 2-1 (SHEET 3 of 3)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Notes:
- a. Up to three sampling locations within 5 miles and in different sectors will be used as available. In addition, one or more control locations beyond 10 miles will be used.
- b. Commercially or recreationally important fish may be sampled. Clams may be sampled if difficulties are encountered in obtaining sufficient fish samples.
- c. If gamma isotopic analysis is not sensitive enough to meet the Minimum Detectable Concentration (MDC), a separate analysis for 1-131 may be performed.
- d. The composite samples shall be composed of a series of aliquots collected at intervals not exceeding a few hours.
- e. If it is found that river water downstream of the plant is used for drinking, drinking water samples will be collected and analyzed as specified herein.
tv
- f. A survey shall be conducted annually at least 50 river miles downstream of the plant to identify those who use water from the
- 1. Altamaha River for drinking.
TABLE 2-2 (SHEET 1 of 2) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Location Direction Distance (a) Sample Type Number Type (a) (miles) 064 Other Roadside Park WNW 0.8 Direct Rad 101 Indicator Inner Ring N 1.9 Direct Rad 102 Indicator Inner Ring NNE 2.5 Direct Rad 103 Indicator Inner Ring NE 1.8 Airborne Rad Direct Rad 104 Indicator Inner Ring ENE 1.6 Direct Rad 105 Indicator Inner Ring E 3.7 Direct Rad 106 Indicator Inner Ring ESE 1.1 Direct Rad Vegetation 107 Indicator Inner Ring SE 1.2 Airborne Rad Direct Rad lOS Indicator Inner Ring SSE 1.6 Direct Rad 109 Indicator Inner Ring S 0.9 Direct Rad 110 Indicator Inner Ring SSW 1.0 Direct Rad III Indicator Inner Ring SW 0.9 Direct Rad 112 Indicator Inner Ring WSW l.0 Airborne Rad Direct Rad Vegetation 113 Indicator Inner Ring W 1.1 Direct Rad 114 Indicator Inner Ring WNW 1.2 Direct Rad 115 Indicator Inner Ring NW 1.1 Direct Rad 116 Indicator Inner Ring NNW 1.6 Airborne Rad Direct Rad 170 Control Upstream WNW (c) River (b) 172 Indicator Downstream E (c) River (b) 201 Other Outer Ring N 5.0 Direct Rad 202 Other Outer Ring NNE 4.9 Direct Rad 203 Other Outer Ring NE 5.0 Direct Rad 204 Other Outer Ring ENE 5.0 Direct Rad 205 Other Outer Ring_ E 7.2 Direct Rad 206 Other Outer Ring ESE 4.S Direct Rad 207 Other Outer Ring SE 4.3 Direct Rad 208 Other Outer Ring SSE 4.S Direct Rad 209 Other Outer Ring S 4.4 Direct Rad 210 Other Outer Ri~ SSW 4.3 Direct Rad 211 Other Outer Ring SW 4.7 Direct Rad 212 Other Outer Ring WSW 4.4 Direct Rad 213 Other Outer Ring W 4.3 Direct Rad 214 Other Outer Ring WNW 5.4 Direct Rad 215 Other Outer Ring NW 4.4 Direct Rad 216 Other Outer Ring NNW 4.S Direct Rad 301 Other Toombs Central School N 8.0 Direct Rad 304 Control State Prison ENE 11.2 Airborne Rad Direct Rad 304 Control State Prison ENE 10.3 Milk 309 Control Baxley S 10.0 Airborne Rad Substation Direct Rad 416 Control Emergency News NNW 21.0 Direct Rad Center Vegetation 2-5
TABLE 2-2 (SHEET 2 of 2) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Notes:
- a. Direction and distance are determined from the main stack.
- b. River (fish or clams, shoreline sediment, and surface water)
- c. Station 170 is located approximately 0.6 river miles upstream of the intake structure for river water, 1.1 river miles for sediment and clams, and 1.5 river miles for fish.
Station 172 is located approximately 3.0 river miles downstream of the discharge structure for river water, sediment and clams, and 1.7 river miles for fish. The locations from which river water and sediment may be taken can be sharply defined. However, the sampling locations for clams often have to be extended over a wide area to obtain a sufficient quantity. High water adds to the difficulty in obtaining clam samples and may also make an otherwise suitable location for sediment sampling unavailable. A stretch of the river of a few miles or so is generally needed to obtain adequate fish samples. The mile locations given above represent approximations of the locations where samples are collected. 2-6
TABLE 2-3 GROUNDWATER MONITORING LOCATIONS WELL DEPTH (ft) MONlTORING PURPOSE RI 82.9 Confined Aquifer Upgradient R2 82.7 Confined Aquifer Near Diesel Generator Bldg. R3 89.2 Confined Aquifer Near CST-J R4 41 Dilution Line Near River Water Discharge Structure RS 33.6 Between Subsurface Drain Lines Downgradient R6 38.2 Between Subsurface Drain Lines Downgradient NW2A 27 Water Table Near CST-2 Inside of Subsurface Drain NW2B 27 Water Table Outside of Subsurface Drain NW3A 26.S Water Table Inside of Subsurface Drain NW3B 2S.3 Water Table Outside of Subsurface Drain NW4A 27 Water Table Upgradient Inside of Subsurface Drain NWSA 26.7 Water Table Upgradient Inside of Subsurface Drain NWSB 26.3 Water Table Upgradient Outside of Subsurface Drain NW6 27 Water Table Near Diesel Generator Bldg. NW8 23 Water Table Near Diesel Generator Bldg. NW9 26.1 Water Table Downgradient Inside of Subsurface Drain NWIO 26.2 Water Table Near CST-2 T3 18 Water Table Near Turbine Bldg. T7 21.4 Water Table Near Diesel Generator Bldg. TIO 18.8 Water Table Near CST-l TI2 23.2 Water Table Near CST-l TIS 27.4 Water Table Near CST-l PISA* 74.S Confined Aquifer Near Turbine Bldg. P1SB 18 Water Table Near Turbine Bldg. PI7A* 77 Confined Aquifer Near Diesel Generator Bldg. PI7B 14.8 Water Table Near Diesel Generator Bldg. Deep Well I 680 Backup Supply for Potable Water (infrequently used) Deep Well 2 711 Plant Potable Water Supply Deep Well 3 710 Potable Water Supply - Rec. Center, Firing Range, and Garage
- Used for water level only 2-7
~N Radiological Environmental Sampling Locations Indicator Control Additional REMP Stations Near TLD .. '" .. the Plant Other
* *
- TLD & Other ,.
',\'~
""" Figure 2-1 2-8
t
)J MO!ltgOJ11(,t I
/
I Tuo /, / !> l 1351
,I I Wayne
. ~N Radiological Environmental Sampling Locations Indicator Control Additional REMP Stations Beyond TLD A.
*
- Five Miles from the Plant Other * *
- TLD & Other ....", .'.,
Figure 2-2 2-9
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* - - - - -j-r: lN~{5~__~:. >
R 1INW [4A __ -~J
- P15A/P156
,
\
TUR61NE aUI'L0 ING l I wens I-i 18 -'-. .'-_.-.-...". . ,===-- ____~ . . . .... .. J'
... . ....... ... ..-- !!
"gure 2 "
.......... ....... ... /
F Ground water Monitormg ' Locations 2-10
Figure 2-4 Deep Wells VISITOR CEI\JTER \ (SAMPLING POINT) I PROTECTED AREA POTABLE POTABLE WATER O..-WATER CHLORINATI~n TANK BLDG oU ~--- DEEP WELL2 -I ~ 1 DISCHARGE STRUCTURE DEEP WELL 3-2-11
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. Since no naturally occurring radionuclides were found in the plant's effluent releases, only man-made radionuclides are reported as pennitted by ODCM 7.1.2.1 . Results for samples collected at locations other than control or indicator stations are discussed in Section 4 under the particular sample type. 3- J
TABLE 3-1 (SHEET 1 of 4) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Edwin I. Hatch Nuclear Plant, Docket Nos. 50-321 and 50-366 Appling County, Georgia Medium or Type and Total Minimum Indicator Location with the Highest Other Control Pathway Number of Detectable Locations Annual Mean Stations(g) Locations Sampled Analyses Concentration Mean (b), Mean (b), Mean (b), (Unit of Performed (MOC) (a) Range Name Distance Mean (b), Range Range Measurement) (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 23.7 Station 107 24.1 NA 22.7 Particulates 312 5.8-50.5 Inner Ring 10.5-50.5 8.1-49.5 (fCi/m3) " -'- '--'-" '~-- '-" ~ -- .. '~~_"' M _W "_ G97120 7t 1.2 miles SE
.. *_..****.....-- -_ .............. _... _._ ....__ .. ...._... (49/49) _
............ ...... ............... __.- ........- --. ..(I~~lJ.Q~}___ .___
Gamma Isotopic 24 Cs-134 50 NOM (c) NOM NOM Cs-137 60 NOM NOM NOM Airborne 1-131 70 NOM NOM NA NOM Radioiodine 318 w (fCi/m3) N Direct Radiation Gamma Dose NA (d) 14.4 Station 104 19.5 14.0 13.6 (mRl91 days) 138 10.4-22.7 Inner Ring 17.3-22.0 9. I -23.5 10.4-19.8 (62/62) 1.6 miles ENE (3 /3) (72172) (12/ 12) Milk Gamma Isotopic (pCi/ l) 24 Cs-134 15 NA NOM NA NOM Cs-\37 18 NA NOM NOM Ba-140 60 NA NOM NOM
.. _-_._
La-140
....... _ **** _ _ w*w.* _ _
- _ _ ** *******
- _ . _ _ _
- _ _ _ * *****
- IS NA
.. _......._...........**_.. _--_._-_._ ... ..... ..-.- .. ... _._-_....... _.. _....... ._-..... NOM -_.__ ....._.... _._.._ .-._-.......... - ..._----'.'" NOM ........_ .-.. .. ..... __ ._-
1-131 1 NA NOM NOM 24
TABLE 3-1 (SHEET 2 of 4) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Edwin I. Hatch Nuclear Plant, Docket Nos. 50-321 and 50-366 Appling County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Number Detectable Locations Annual Mean Locations Sampled of Analyses Concentration Mean (b), Mean (b), (Unit of Performed (MOC) (a) Range Name Distance Mean (b), Range Measurement) (Fraction) & Direction Range (Fraction) (Fraction) Vegetation Gamma (pCilkg-wet) Isotopic 36 1-131 60 NDM NOM NOM Cs-134 60 NOM NOM NOM Cs-l37 80 44 .3* Station 106 48 .3* 28.99 15 .0-84.0 Inner Ring 15.0-84.0 (1112) (9/24) 1.1 miles ESE (8/12) River Water Gamma (pCi/l) Isotopic w , 26 w Mn-S4 15 NOM NOM NOM Fe-59 30 NOM NOM NOM Co-58 15 NOM NOM NOM Co-60 15 NOM NOM NOM Zn-65 30 NOM NOM NOM Zr-9S 30 NOM NOM NOM Nb-9S 15 NOM NOM NOM 1-131 15 (e) NOM NOM NOM Cs-134 15 NOM NOM NOM Cs-J37 18 NOM NOM NOM Ba-140 60 NOM NOM NOM La-140 15 ......... __ ._-_......... _..* NOM NOM NOM Tritium 3000 (f) * * **T93 - ..........__._......_-_.. .__... ----_ ...... __ ... _--------- --- Station 172
- ** M ** *** **** * * * ***
364
. fi;g--- -
..... . ...*.....*. --...- --.-....
8 (114) 3.0 miles (1/4) (114) Downstream
TABLE 3-1 (SHEET 3 of 4) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Edwin I. Hatch Nuclear Plant, Docket Nos. 50-321 and 50-366 Appling County, Georgia Medium or Type and Minimum Indicator Location with the Highest Control Pathway Total Number Detectable Locations Annual Mean Locations Sampled of Analyses Concentration Mean (b), Mean (b), (Unit of Performed (MDC) (a) Range Name Distance Mean (b), Range Measurement) (Fraction) & Direction Range (Fraction) (Fraction) Fish Gamma (pCi/kg-wet) Isotopic 8 Mn-54 130 NDM NDM NDM Fe-59 260 NDM NDM NDM Co-58 130 NDM NDM NDM Co-60 130 NDM NOM NOM Zn-65 260 NDM NOM NOM Cs-134 130 NDM NOM NOM Cs-137 150 NOM NOM NOM Sediment Gamma (pCi/kg-dry) Isotopic 4 Cs-134 150 NDM NOM NOM Cs-137 180 30.3 Station 172 91.7 91.7 (1/2) 3.0 miles (1 /2) (1 /2) Downstream
TABLE 3-1 (SHEET 4 of 4) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Edwin I. Hatch Nuclear Plant, Docket Nos. 50-321 and 50-366 Appling County, Georgia NOTATIONS
- 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 specified locations that are detectable is placed in parenthesis.
- c. No Detectable Measurement(s).
w, Vl
- d. Not Applicable.
- e. If a drinking water pathway were to exist, a MDC of 1 pCi/1 would have been used (see Table 4-1 of this report).
- f. If a drinking water pathway were to exist, a MDC of 2000 pCi/1 would have been used (see Table 4-1 of this report).
- g. "Other" stations, identified in the "station type" column of Table 2-2, include community and special stations.
4.0 DISCUSSION OF RESULTS Included in this section are evaluations of the laboratory results for the various sample types. The Minimum Detectable Difference (MOD) compares the lowest significant difference between a control station and an indicator station, or the control station and the community station, that can be determined statistically at the 99% Confidence Level (CL). The MOD was determined using the standard Student's t-test. MOD as a tool can quantify plant Hatch's impact on the surrounding environment. A difference in the mean values which was less than the MOD was considered to be statistically indiscernible. The 2012 results were compared with past results, including those obtained during pre-operation. 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 analyses. 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 (NOM) were observed or periods for which values were not applicable (e .g., milk indicator, etc.) are plotted as O's and listed in the tables as NOM. Table 4-1 Minimum Detectable Concentrations (MDC) Analysis Water Airborne Fish Milk Grass or Sediment (pCi/I) Particulate (pCi/kg- (pCi/I) Leafy (pCi/kg-or Gases wet) Vegetation dry) (fCUm3) (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 IS 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 I (bJ 70 I 60 Cs-I34 IS 50 130 15 60 150 Cs-I37 18 60 150 18 80 180 Ba-I40 60 60 La-140 15 IS (a) If no drinking water pathway exists, a value of 3000 pCi/1 may be used . (b) If no drinking water pathway exists, a value of 15 pCi/1 may be used . 4-1
Table 4-2 Reporting Levels (RL) Analysis Water Airborne Fish Milk (pCi/l) Grass or Leafy (pCi/I) Particulate (pCi/kg-wet) Vegetation or Gases (pCi/kg-wet) (fCi/m3) H-3 20,000 (a) Mn-54 1000 30000 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-140 100 400 (a) This is the 40 CFR 141 value for drinking water samples. If no drinking water pathway exists, a value of 30,000 may be used. (b) Ifno drinking water pathway exists, a value of20 pCill may be used. 4-2
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-13 7, continue to be detectable . Significant upward trends also followed the Chernobyl incident which began on April 26, 1986. The most significant nuclear event since Chernobyl occurred at Fukushima Daiichi Nuclear Power Plant after the Tohoku earthquake and tsunami on March 11, 201l. Equipment failures and nuclear meltdowns resulted in radioactivity being released into the atmosphere. Southern Nuclear's three sites (Farley, Hatch, and Vogtle) detected 1-131 in REMP samples for several weeks following the disaster. 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. All results were tested for conformance to 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. One data point was excluded from the data set for failing Chauvenet's criterion . This was an abnormally high value for Cs-13 7 in a forage sample. Data exclusions are discussed in this section under the appropriate sample type. 4-3
TABLE 4-3 DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLES 151 Quarter OSL Dosimeters No data collected Badges were missing in the field at Replaced OSL dosimeters CR 632212/TE 633427 Station # 113 collection time at beginning of quarter W 1.1 miles W 05/07/12-05114112 AC/AF No sample collected due Electrical power outage Sampling resumed CR 625129/TE 628992 Station #304 to power outage following power restoration 06/11/12-06/18/12 AC/AF MDC could not be met Pump was found off Pump was replaced CR 625133/TE 628987 Station # 116 due to low sample volume 09/03/12-09/10/12 AC/AF No sample collected due Electrical power outage Sampling resumed CR 625136ITE 628994 Station # 107 to power outage following power restoration 09/10112-09/17/12 AC/AF No sample collected due Electrical power outage Sampling resumed CR 625136ITE 628994 Station # 107 to power outage following power restoration 09/17112-09/24/12 AC/AF No sample collected due Electrical power outage Sampling resumed CR 625136ITE 628994 Station # I 07 to power outage following power restoration 4th Quarter OSL Dosimeters No data collected Badges were missing in the field at Replaced OSL dosimeters CR 632212/TE 633427 Station # 104 collection time at beginning of quarter W 1.6 miles ENE
4.1 Land Use Census and River Survey In accordance with ODeM 4.1 .2, a land use census was conducted on November 26, 2012, to detennine the locations of the nearest permanent residence and milk animal in each of the 16 compass sectors within a distance of 5 miles, and the locations of all milk animals within a distance of 3 miles. A milk animal is defined as a cow or goat producing milk for human consumption. The locations of beef cattle and of gardens greater than 500 square feet producing broad leaf vegetation were also included in the census. The census results are tabulated in the Table 4 . 1-l. Table 4.1-1 LAND USE CENSUS RESULTS Distance in Miles to Nearest Location in Each Sector SECTOR RESIDENCE MILK ANIMAL BEEF CATTLE GARDEN N 2 .8 None 4.7 3.6 NNE 2.9 None None 4.7 NE 3.3 None 3.4 4.8 ENE 4.2 None 4.1 4.6 E 3.0 None None None ESE 3.8 None None None SE l.8 None 2.4 4.4 SSE 2.0 None 3.6 2.1 S l.1 None 2.5 l.5 SSW l.3 None 2.0 3.6 SW l.1 None 2.3 l.6 WSW l.0 None 3.6 2.0 W l.1 None 2.7 None WNW l.1 None None None NW 3.6 None 4.5 3.7 NNW 1.8 None 2.8 2.8 4-5
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. No change in the controlling receptor was required as a result of the 2012 land use census. The current controlling receptor as described in ODCM Table 3-7 is a child in the WSW Sector at 1.0 miles. ODCM 4.1.2.2.2 requires that whenever the land use census identifies a location which would yield 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). The 2012 land use census did not identify a garden which yielded a calculated dose 20% greater than that for any of the current indicator stations for vegetation. As required by Note f of Table 2-1, the annual survey of the Altamaha River for approximately 50 miles downstream of the plant was conducted on December 10, 2012 to identify any withdrawal of river water for drinking purposes. No sources of withdrawal for drinking water were identified. Information obtained from the Georgia Department of Natural Resources in September 2012 indicated that no surface water withdrawal permits for agricultural or drinking purposes had been issued for this stretch of the Altamaha River between the 2011 survey and the 2012 survey. Should it be determined that river water downstream of the plant is being used for drinking, the sampling and analysis requirements for drinking water found in Table 2-1 would be implemented. Irrigation equipment was identified at Clarke's Farm about 3i4 mile downstream of Station #172 river water sampling station. The equipment is potentially used to irrigate peanuts. Mr. Clarke was contacted in September and November of 2012 and he stated that he has not irrigated his peanut crop from the river in 2012. Should it be determined that river water downstream of the plant is being used for irrigation, additional sampling and analysis of the crop would be implemented. 4-6
4.2 Airborne As indicated in Table 2-2 and Figures 2-1 and 2-2, airborne particulates and airborne radioiodine are collected at 4 indicator stations (Nos. 103, 107, 112 and 116) which encircle the plant near the site periphery and at 2 control stations (Nos. 304 and 309) which are located approximately 10 miles from the main stack. At each location, air is continuously drawn through a glass fiber filter and a charcoal canister placed in series to collect airborne particulates and radioiodine. The filters and canisters are collected weekly and analyzed for gross beta and 1-131, respectively. A gamma isotopic analysis is performed quarterly on a composite of the filters for each station. The 2012 annual average weekly gross beta concentration of 23.7 fei/m 3 for the 3 indicator stations was 1.0 fei/m more than that for the control stations (22.7 3 fei/m ). This difference is not statistically discernible, since it is less than the 3 calculated MOD of 2.5 fei/m . Figure 4.2-1 and Table 4.2-1 provide the historical trending of the average weekly gross beta concentrations in air. In general, there is close agreement between the results for the indicator and control stations. This close agreement supports the position that the plant is not contributing significantly to the gross beta concentration in air. Figure 4.2-1 Average Weekly Gross Beta Air Concentration 300 . I, I II I
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I i o Po 75 77 79 81 83 85 87 89 91 93 95 97 99 3 5 7 9 11 Year
- - MOC _ Indicator - . - Control I 4-7
Table 4.2-1 A verage W ee klIy G ross B et a A*Ir Concen t ratIOn Year Indicator Control (fCi/m3) (fCi/m3) Pre-op 140 140 1974 87 90 1975 85 90 1976 135 139 1977 239 247 1978 130 137 1979 38 39 1980 49 48 1981 191 203 1982 33 34 1983 31 30 1984 26 28 1985 22 21 1986 36 38 1987 23 22 1988 22.6 2l.7 1989 18.4 17.8 1990 19.3 IS.7 1991 18.1 18 1992 18.5 18.4 1993 20.4 20.7 1994 19.5 19.7 1995 21.7 2l.7 1996 21.3 21.4 1997 20.3 20.7 1998 20.0 20.5 1999 21.3 21.3 2000 23.6 23.9 2001 21.5 2l.0 2002 19.3 19.2 2003 18.8 18.2 2004 21.4 21.3 2005 19.7 19.4 2006 24.9 24.7 2007 24.4 24.3 2008 21.8 22.5 2009 2l.2 21.4 2010 23.1 24.0 2011 23.5 25.1 4-8
Table 4.2-1 (continued) A verage W ee klIy G ross Be ta A'Ir C oncen t ra f IOn Year Indicator Control (fCi/m3) (fCi/m3) 2012 23.7 22.7 During 2012, no man-made radionuclides were detected from the gamma isotopic analysis of the quarterly composites of the particulate air filters. During preoperation and during operation through 1986, a number of fission products and activation products were detected. These were generally attributed to the nuclear weapons tests and to the Chemobyl incident. On only one occasion since 1986, has a man-made radionuclide been detected in a quarterly composite. A small amount of Cs-l37 (1.7 fCi/m3) was identified in the first quarter of 1991 at Station 304. The MDC and RL for Cs-l37 in air are 60 and 20,000 fCi/m3, respectively. The historical trending of the average annual concentrations of detectable Cs-137 from quarterly air filter composites is provided in Figure 4.2-2 and Table 4.2-2. Figure 4.2-2 60 r,_.._- Average Annual Cs-137 Concentration in Air
..- .._ ' - .. ,---,--,.. ...- r *- - .... , - - _.
50
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- 1 ....ra c:::
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0 I~ -+- -+ I:P -,- ~ II"i't ~ - I~ ~'"' Po 75 77 79 81 83 85 87 89 91 93 95 97 99 3 5 7 9 11 Year
--+--Indlcator - - - Control - - MDC I 4-9
Table 4.2-2 A verage A nnua Ie s- 137 C oncentratIOn I n A"Ir Year Indicator Control (fCilm3) (fCi/m3) Pre-op NDM 2.0 1974 l.5 2.0 1975 1.4 1.4 1976 0.6 0.7 1977 l.5 1.4 1978 2.3 2.6 1979 0.8 0.8 1980 0.4 0.6 1981 l.8 l.7 1982 0.5 0.6 1983 0.7 NDM 1984 NDM NDM 1985 0.7 NDM 1986 8.1 9.6 1987 NDM NDM 1988 NDM NDM 1989 NDM NDM 1990 NDM NDM 1991 NDM l.7 1992 NDM NDM 1993 NDM NDM 1994 NDM NDM 1995 NDM NDM 1996 NDM NDM 1997 NDM NDM 1998 NDM NDM 1999 NDM NDM 2000 NDM NDM 2001 NDM NDM 2002 NDM NDM 2003 NDM NDM 2004 NDM NDM 2005 NDM NDM 2006 NDM NDM 2007 NDM NDM 2008 NDM NDM 2009 NDM NDM 2010 NDM NDM 2011 NDM NDM 4-10
Table 4.2-2 (continued) A verage A nnuaIe s- 137 Concen t ra f IOn I n A*Ir Year Indicator Control (fCiim3) (fCii m3) 2012 NDM NDM During 1976, 1977, and 1978, positive levels of 1-131 were found in nearly all of the samples collected for a period of a few weeks following atmospheric nuclear 3 weapons tests. Some of the concentrations were approximately 70 fCi i m . In 1986, the same phenomenon occurred following the Chernobyl incident. The nuclear accident at Fukushima Daiichi Nuclear Power Plant which occurred after the Tohoku earthquake and tsunami on March 11, 20 II released radioactivity into the environment that was detected in Hatch air samples. Iodine-131 was detected in air cartridges after Fukushima but no changes in gross beta activity were seen during that same time period. Iodine-131 ranging from 16.9-106.9 fCi /m 3 was seen at Hatch for several weeks following the Fukushima accident. The highest airborne 1-131 concentration found to date in an individual charcoal canister was 3 3 217 fCi/m in 1977. The MDC and RL for airborne 1-131 are 70 fCi/m and 900 3 fCi /m , respectively. Table 4-3 lists REMP deviations that occurred in 2012. There were four air sampling deviations due to power supply issues (Station #304 in May was weather related, and Station # 107 in September - three different sample collection time periods -all weather related). There was one sample deviation at Station 116 due to pump failure. 4-\\
4.3 Direct Radiation In 2012, direct (external) radiation was measured with Landauer InLight optically stimulated luminescent (OSL) dosimeters . The OSL dosimeters replaced Panasonic thermoluminescent dosimeters (TLOs). The Panasonic system was retired at the end of 2010 due to the inability to keep the aging badge readers operating reliably. Similar to the TLD protocol of the past, two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges . Two direct radiation stations are establ ished in each of the 16 compass sectors around the plant to form 2 concentric rings, as seen in Figures 2-1 and 2-2. The two ring configuration of stations was established in 1980, in accordance with NRC Branch Technical Position "An Acceptable Radiological Environmental Monitoring Program", Revision I, 1979. With the exception of the East sector, the inner ring stations (Nos. 101 through 116) are located near the site boundary and the outer ring stations (Nos. 201 through 216) are located at distances of 4 to 5 miles from the plant. The stations in the East sector are a few miles farther out than the other stations in their respective rings due to large swamps making normal access extremely difficult. The 16 stations forming the inner ring are designated as the indicator stations. The 3 control stations (Nos. 304, 309 and 416) are located 10 miles or more from the plant. Stations 064 and 30 I monitor special interest areas. Station 064 is located at the onsite roadside park, while Station 30 I is located near the Toombs Central School. Station 210, in the outer ring, is located near the Altamaha School (the only other nearby school). As provided in Table 3-1, the average quarterly exposure measured at the indicator stations (inner ring) during 2012 was 14.4 mR. At the control stations , the average quarterly exposure was 13.6 mR. This difference (0.8 mR) is not statistically discernible since it is less than the MOD of 2.5 mR. The quarterly exposures acquired at the outer ring stations during 2012 ranged from 9.1 to 23.5 mR, with an average of 14.1 mR. The average for the outer ring stations was 0.5 mR more than the average for the control stations. Since the results for the outer ring stations and the control stations differ by less than the MOD of 2.4 mR, there is no discernible difference between outer ring and control station results for 2012. 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 TLOs from Teledyne to Panasonic. It should be noted however that the differences between indicator and control and outer ring values did not change. 4-12
During 2010, OSL badges were co-located on station with the TLD badges. In 2011, only the OSL badges were placed at each station. Following the change to only OSL badges, the differences between indicator, control, and community locations has been consistent with previous years. An increase noted in 2010 reflects issues (especially during 2 n Qtr) with the aging Panasonic TLD reader. The close agreement between the station groups supports the position that the plant is not contributing significantly to direct radiation in the environment. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 25 V v I Po 75 77 79 81 83 85 87 89 91 93 95 97 99 01 03 05 07 09 11 Year
--Indicator ____ Control -,l- Outer Ring 4-13
Table 4.3-1 Average Quarterly Exposure from Direct Radiation Year Indicator (mR) Control (mR) Outer Ring (mR) Pre-op 22.3 23 NA 1974 23 .2 25.6 NA 1975 10.0 10.5 NA 1976 8. 18 6.9 NA 1977 7.31 6.52 NA 1978 6.67 6.01 NA 1979 5. 16 6.77 NA 1980 4.44 5.04 4.42 1981 5.9 5.7 5.7 1982 12.3 12 11.3 1983 11.4 11.3 10.6 1984 13.3 12.9 11. 9 1985 14.7 14.7 13.7 1986 15 14 14.5 1987 14.9 14.6 15.3 1988 15 .0 14.7 15.2 1989 16.4 18.0 16.5 1990 14.9 13.9 14.7 1991 15. 1 13.7 15 .6 1992 11.9 10.9 12.3 1993 11.6 10.7 11.5 1994 11 10.7 11.2 1995 11.5 10.8 11.3 1996 11.6 11.3 11.6 1997 12.3 11.8 12.3 1998 12.1 12.3 12.3 1999 12.8 13.2 13 .0 2000 13 .6 13.3 13 .3 2001 12.0 12.1 11.8 2002 11.7 11.7 11.5 2003 11.4 11.4 11.4 2004 12.2 12.4 12.2 2005 12. 1 12 .5 12.0 2006 12.4 11.9 11.8 2007 12.8 12.5 12.6 2008 13 .0 12.3 12.4 2009 12.4 12.2 12.2 2010 15 .8 15.6 16.0 2011 13.7 13.1 13.1 4-14
Table 4.3-1 (continued) Average Quarterly Exposure from Direct Radiation Year Indicator (mR) Control (mR) Outer rung (mR) 2012 14.4 13.6 14.1 The historical trending, since 1986, of the average quarterly exposures at the special interest areas is 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 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 25 20 I ~~
~ 15
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o D 10 5 o 86 87 88 89 90 91 92 93 94 95 96 97 98 99 0 1 02 3 4 5 6 7 8 9 10 11 12 Year
- + - Roadside Park (Sta 064) ___ Toombs Central School (Sta 301) 4-15
Table 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas Period Station 064 Station 301 (mR) (mR) 1986 14.6 15 . 1 1987 14.2 15.0 1988 14.9 15.3 1989 16. 1 16.6 1990 15.1 14.4 1991 14.4 15 .2 1992 1 1.1 11.5 1993 11.2 10.8 1994 10.4 10.7 1995 11.0 10.5 1996 11. 7 11.0 1997 12.6 11.4 1998 12.4 11.8 1999 12.5 12.4 2000 13.3 12.6 2001 11.8 11.3 2002 11.4 11.4 2003 11.2 11.1 2004 11.9 12.3 2005 11.8 12.4 2006 11.9 11.6 2007 1l.9 12.1 2008 12.3 12.2 2009 12. 1 12.1 2010 15.7 15 .5 2011 12.7 13 .1 2012 14.1 13.2 Table 4-3 lists the RE MP Program deviations that occurred in 2012. There was one deviation involving OSL dosimeters. At first quarter collection, badges from Station # 1 13 were missing in the field at the time of collection. 4-16
The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self imposed limit of 3.S. Previously with TLOs, this limit had been lA. However, the OSL readings varied more (between the two elements) than the TLO readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication ISO, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 20 I O. The limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater or equal to 3.S are excluded from the data set since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 2012, the following OSL results were excluded from the data set because their standard deviations were greater than or equal to 3.S : First Quarter None Second Quarter None Third Quarter None Fourth Quarter H416B If one badge at a station exhibited a standard deviation greater than or equal to 3.S, then the reading of the companion badge at each location would be used to determine the quarterly exposure. 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 conection factors were reviewed. No reason was found for the high standard deviations. A major advantage of the OSL badge is that it can be read multiple times. A new practice was employed in 2011 to re-read any environmental badges that yielded a standard deviation > 3.S. The readings with the lower standard deviation would be reported. 4-17
4.4 Milk Milk samples are obtained biweekly from Station 304 (the state prison dairy) which is a control station located more than 10 miles from the p1ant. Gamma isotopic and 1-131 analyses are performed on each sample as specified in Tables 2-1 and 2-2. Since 1989, efforts to locate a reliable milk sample source within 5 miles of the plant have been unsuccessful. During 2012, no man-made radionucl ides were detected from the gamma isotopic analysis of the milk samples. Cesium-137 was found in most of the samples each year from 1978 (when this analysis became a requirement) through 1989. No other man-made radionuclides have been detected by this analysis. The MDC and RL for Cs-137 in milk are 18 and 70 pC il l, respectively. The historical trending of the average annual detectable Cs-137 concentration in milk is provided in Figure 4.4-1 and Table 4.4-1. Figure 4.4-1 Average Annual Cs-137 Concentration in Milk 25 20 - U -Q. c:: 15 \ ~
/'. V
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0
-... lU ~ r-' 1\ r c:: Q) 10 ~
\
\V'\ \ I
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Po 75 77 79 81 83 85 87 89 91 93 95 97 99 01 03 05 07 09 11 Year
-Indicator - - - Control - - MOC I 4-18
Table 4.4-1 A verage A nnua I C s- 137 C oncen t ra f.on III "M"lk I Year Indicator Control (pCill) (pCi/l) Pre-op 19.9 19.4 1974 NOM NOM 1975 NOM NOM 1976 NOM NOM 1977 NOM NOM 1978 12.1 18.3 1979 16.1 13 1980 14.7 15.4 1981 12.57 10.2 1982 11.8 11 1983 12 7.2 1984 9.6 10.2 1985 9.14 5.35 1986 9.8 10 1987 NOM NOM 1988 10.9 NOM 1989 8.6 7.9 1990 NOM NOM 1991 NOM NOM 1992 NOM NOM 1993 NOM NOM 1994 NOM NOM 1995 NOM NOM 1996 NOM NOM 1997 NOM NOM 1998 NOM NOM 1999 NOM NOM 2000 NOM NOM 2001 NOM NOM 2002 NOM NOM 2003 NOM NOM 2004 NOM NOM 2005 NOM NOM 2006 NOM NOM 2007 NOM NOM 2008 NOM NOM 2009 NOM NOM 2010 NOM NOM 2011 NOM NOM 4-19
Table 4.4-1 (continued) A verage A nnua Ie s- 137 C oncen t ra f Ion III , M'lkI Year Indicator Control (pCi/1) (pCi/I) 2012 NDM NDM During 2012, 1-131 was not detected in any of the milk samples. During preoperation, all readings were less than 2 pCi/1 which was the allowed MDC at that time. Figure 4.4-2 and Table 4.4-2 provide the historical trending of the average annual detectable concentration of 1-131 in milk. In 1988, a single reading of 0.32 pC il l, which was believed to have resulted from a procedural deficiency, was reported . The MDC and RL for 1-131 in milk are I and 3 pCi/ l, respectively. All the detectable results for Cs-137 and 1-131 are attributed to fallout from the nuclear weapons tests and the Chernobyl incident. Figure 4.4-2 Average Annual 1-131 Concentration in Milk 16 14
/
-S U 12 - a. 10 ......... c: 0 1\1 8 c: Q) u c: 6
\\
0
..
,
>
c.> 4 2 ~\ I
/ l\ ~
0 V/ N 'It' ~ Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
-+-- Indicator ----- Control - MOC RL I 4-20
Table 4.4-2 A verage A nnua -131 Concen tra f.on m I I . M*lk I Year Indicator Control (pCiJl) (pCiJl) Pre-op NOM NOM 1974 0.98 2.6 1975 0.3 NOM 1976 12.23 9.1 1977 14.61 4.08 1978 2.72 4.18 1979 NOM NOM 1980 l.26 0.69 1981 NOM NOM 1982 NOM NOM 1983 NOM NOM 1984 NOM NOM 1985 NOM NOM 1986 8.9 7.6 1987 NOM NOM 1988 NOM 0.32 1989 NOM NOM 1990 NOM NOM 1991 NOM NOM 1992 NOM NOM 1993 NOM NOM 1994 NOM NOM 1995 NOM NOM 1996 NOM NOM 1997 NOM NOM 1998 NOM NOM 1999 NOM NOM 2000 NOM NDM 2001 NOM NOM 2002 NOM NOM 2003 NOM NOM 2004 NOM NOM 2005 NOM NOM 2006 NOM NOM 2007 NOM NOM 2008 NOM NOM 2009 NOM NOM 2010 NOM NOM 2011 NOM NOM 4-21
Table 4.4-2 (continued) A verage A nnua I I-131 C oncen tra f IOn III " M"]k I Year Indicator Control (pCill) (pCi/l) 2012 NDM NOM 4.5 Vegetation In accordance with Tables 2-1 and 2-2, grass samples are collected monthly from two indicator stations near the site boundary (Nos. 106 and 112) and at one control station located about 21 miles from the plant (No. 416). Gamma isotopic analyses are performed on each sample. Gamma isotopic analysis on vegetation samples began in 1978 when the analysis became a TS requirement. The results presented in Table 3-1 show that Cs-137 was the only man-made radionuclide detected in vegetation samples during 2012. Cesium-137 was detected in 9 samples of the 24 samples collected at the indicator stations. One sample result was excluded because it did not pass Chauvenet's criterion. No reason for the abnonnally high result could be determined. The average of the remaining 8 samples was 44 .3 pCi/kg-wet. Cesium-137 was detected in one control station sample at 29 .0 pCi/kg-wet. Due to the low number of samples, MOD was not able to be used to evaluate the data. In 2011, the average Cs-137 seen at the indicator station was higher than it had been since the late 1980s. Fertilization of the area, resulting in soil disturbance could have accounted for the increase. However, the Cs-137 detected at indicator stations could potentially be attributed to plant effluents. In 2012, the Cs-137 values returned to historical range. Since 1986, Cs-137 has been the only man-made radionuclide found in vegetation samples. The MOC and RL for Cs-137 in vegetation samples are 80 pCi/kg-wet and 2000 pCi/kg-wet, respectively. The occasional presence of Cs-137 in vegetation samples is attributed primarily to fallout from nuclear weapons tests and the Chernobyl incident. Figure 4.5-1 and Table 4.5-1 provide the historical trending of the average annual detectable Cs-137 concentration found in vegetation. Since 1978, the Cs-137 concentration has been on a decline, and since about 1989, generally occurring below the required MDC. In March 2011, after the nuclear accident at Fukushima Oaiichi Nuclear Power Plant, Southern Nuclear's three sites (Farley, Hatch, and Vogtle) detected I-131 in REMP samples for several weeks following the disaster. Iodine-131 was detected at Hatch in two of the three forage samples collected on 03 /28/11 (after the Fukushima event), but not in any forage samples collected since that time. The range of I-131 values was 85.4 to 90.1 pCi/kg-wet. The MDC and RL for I-131 in vegetation are 60 and 100 pCi/kg-wet, respectively. 4-22
Figu re 4.5-1 Average Annual Cs-137 Concentration in Vegetation 1200
,
-- (I) 1000
~I 1\ )
- U Cl
.::t:. 800
/
- a. c: 600 I --...
.~
ra c(I): 400 ~
/ 1\ / \.
() c: 0 (.) I 200 I '- ~/ N/ :\ .,... I\ 0 ~ V r-4 ~ ~ ... ~r ,""" H Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
-Indicator - - - Control - - MOC 4-23
Table 4.5-1 A verage A nnua I C s- 137 C oneen t ra f IOn III
. Vege t a f IOn Year Indicator Control (pCilkg-wet) (pCi/kg-wet)
Pre-op 55 30 1974 NOM NOM 1975 NOM NOM 1976 NDM NOM 1977 NOM NDM 1978 112 1089 1979 59 695 1980 208 916 1981 182 152 1982 65 99 1983 95 211 1984 149 388 1985 60.9 113.3 1986 80 215 1987 60 428 1988 40.1 228.8 1989 37 NOM 1990 66.7 34.5 1991 34.1 36.1 1992 35.2 41.3 1993 24.7 45.8 1994 32.2 46.6 1995 49 .8 47.6 1996 47 .2 41.1 1997 48.4 54.9 1998 81.4 44.1 1999 26.9 NDM 2000 NOM NOM 2001 NOM NOM 2002 33 .7 41.1 2003 61.0 62.8 2004 41.6 43 .5 2005 47 .7 39.8 2006 66.8 29.6 2007 55.7 31.1 2008 41.8 38.1 2009 46.8 NDM 2010 31.4 NOM 2011 267.5 19.1 4-24
Table 4.5-1 (continued) A verage A nnua Ie s- 137 C oncen tra fIOn In v ege ta f IOn Year Indicator Control (pCi/kg-wet) (pCi/kg-wet) 2012 44.3 29.0 4.6 River Water Surface water from the Altamaha River is obtained at an upstream location (Station 170) and at a downstream location (Station 172) using automatic samplers. Small quantities are drawn at intervals not exceeding a few hours. The samples drawn are collected monthly and quarterly composites are produced from the monthly collections. As specified in Table 2-1, a gamma isotopic analysis is conducted on each monthly sample. No man-made gamma emitting nuclides were detected during 2012. The only man-made gamma emitters previously detected are presented in the table below. Year Quarter Station Radionuclide Level (pCi/J) 1975 4th 172 Ce-141 78.2 1986 2nd 170 La-140 18.0 1986 2nd 172 Cs-137 12.0 1988 2nd 170 Cs-137 6.8 A tntlUm analysis is performed on the quarterly composite. Prior to 1986, positive results were usually found in each quarterly composite at levels generally ranging from 200 and 350 pCi/1 which is approximately background environmental levels. Subsequently, the number of positive results have diminished. In 2012 quarterly samples, tntlum was detected in one of the four quarterly samples at the upstream (control) location at 195 pCi/l, and in one of the four quarterly samples at the downstream (indicator) location at 364 pCi/1. Due to the low number of samples, MOD was not able to be used to evaluate the data. The low levels detected at both the indicator and control stations are consistent with detectable values observed in past samples. The MOC and RL for tritium in river water are 3000 and 30,000 pCi/l, respectively. Figure 4.6-1 and Table 4.6-1 provide the historical trending of the annual average detectable tritium concentration in river water. The annual downstream survey of the Altamaha River to determine if river water is being withdrawn for drinking purposes is discussed in Section 4.1. 4-25
Figure 4.6-1 Average Annual H-3 Concentration in River Water 3500 3000 S 2500 - U Cl. ..... c 2000 o 10 1: 1500 CIJ u c <3 1000
,
.,
500 l l/. :---.. ~~ II. o It
~
.J~
~ ~~ V~
" ./. ~ / tJK I 1~
Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
-+-Indicator - Control - - MOC 4-26
Table 4.6-1 A verage A nnua IH3C
- . ru ver Wt oncen t ra f IOn In a er Year Indicator Control (pCi/I) (pCi/l)
Pre-op 210 191 1974 230 205 1975 205 238 1976 165 153 1977 189 170 1978 224 193 1979 210 180 1980 358 218 1981 220 135 1982 165 220 1983 265 328 1984 437 327 1985 288 220 1986 242 206 1987 241 204 1988 220 NOM 1989 NOM NOM 1990 139 NOM 1991 NOM NOM 1992 NOM NOM 1993 NOM NDM 1994 NOM NDM 1995 200 NOM 1996 144 147 1997 NDM NOM 1998 NOM NOM 1999 NOM NOM 2000 209 NOM 2001 NOM NOM 2002 NOM NOM 2003 NOM 261 2004 206 302 2005 245 NOM 2006 299 NOM 2007 235 338 2008 329 298 2009 242 343 2010 403 426 2011 366 293 4-27
Table 4.6-1 (continued) A A verage nnua lH3C - oncen t ra f IOn ID. ill ver Wt a er Year Indicator Control (PCi/l) (pCill) 2012 195 364 4.7 Fish Gamma isotopic analyses were perfonned on the edible portion of the fish samples collected at the river stations on April 9, 2012 and November I, 2012. The control station (No. 170) is located upstream of the plant while the indicator station (No. 172) is located downstream. As shown in Table 3-1, no man-made radionuclides were detected in fish during 2012 . Cs-137 in past fish samples has been attributed primarily to weapons testing and the Chemobyl incident. However, the Cs-137 seen in past fish samples at the indicator station could potentially be attributed to plant effluents. The MDC and RL for Cs-13 7 in fish are ISO and 2000 pCi/kg-wet, respectively. The historical trending of the average annual detectable Cs-137 concentration in fish is provided in Figure 4.7-1 and Table 4.7-1. Figure 4.7-1 indicates, in general, a decline in the Cs-137 levels after 1983. (Note: From 1979 through 1982, clams were collected rather than fish.) 4-28
Figure 4.7-1 Average Annual Cs-137 Concentration in Fish 160 I I 140 - Q)
~I 120 .' III \
-- U Ol
..lI::
100 I I\ \ 1\ .3: c: \ t-t r-4
.I II/ \
-- 80
~~
V ~ 0
\
I1l
....
c: 60 . j
;--
- t-- \. JI II I r-- ~ t?
1\V\
~ ~l -
Q)
\
(J j
.. ~-... -~
c: 40 i--- 0 U 1\ / r-,
~
~P ~
~N ~
20
'" .,~
~~
~ "\ ~ tt ~ II. / ~h 0
Po 75 77 79 81 83 85 87 89 91 93 95 97 99
. ,
03 05 07
-..~
09
-- ~
11 Year
-+-Indicator
-Control 4-29
Table 4.7-1 A verage A nnua Ie s- 137 Concen t ra f JOn 10 "F"IS h Year Indicator Control (pCi/kg-wet) (pCilkg-wet) Pre-op 90 115 1974 134 61 1975 80.6 89.4 1976 73 88 1977 76 91 1978 88 47 1979 NOM NOM 1980 NOM NOM 1981 NOM NOM 1982 NOM NOM 1983 138.6 67.5 1984 84 53 1985 117 63.3 1986 79 44 1987 62 52 1988 77.8 33.3 1989 34.3 28.9 1990 26.7 24.2 1991 32.9 26.9 1992 41.6 28 .8 1993 38.0 25.9 1994 23.8 20.7 1995 25.0 27.9 1996 20.4 18.0 1997 29.4 15.1 1998 26.l 17.7 1999 22.3 13 .5 2000 17.9 25.3 2001 20.8 10.2 2002 18.2 13.0 2003 13.1 7.1 2004 11.6 18.8 2005 13.0 13.3 2006 10.4 13.5 2007 6.8 9.8 2008 19.9 8.4 2009 12.4 8.4 2010 11.6 8.6 2011 8.6 7.1 4-30
Table 4.7-1 (continued) A verage A nnua . FOIS h l e S- 137 C oncen t ra f IOn III Year Indicator Control (pCi/kg-wet) (pCi/kg-wet) 2012 NDM NDM In the past, the only other man-made radionuclides detected in fish samples were Co-60 and Cs-134. During pre-operation, Co-60 was detected in one fish sample at a very low concentration. During the period of 1983 through 1988, Cs-134 was found in about half of the samples at concentrations of the same order of magnitude as those found for Cs-137 . The Co-60 and Cs-134 levels found in these samples are attributed to the nuclear weapons tests and the Chernobyl incident. Figure 4.7-2 and Table 4.7-2 show the historical trending of the annual average detectable concentration of Cs-134 in fish. Figure 4.7-2 Average Annual Cs-134 Concentration in Fish 160 I 140 -OJ 120
~
Cl ..:.::
- 100 t>
-gc. 80 \
...ra
'E 60
\
(1) (J g 40
\
/
\
/\
t> \ 20 "'-. o
/ \ ) '\~
Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
--+-Indicator ----- Control - MOC 4-31
Table 4.7-2 A vera~e A nnua I C s- 134 C oneen tra fIOn .III F*IS h Year Indicator Control (pCilkg-wet) (pCilkg-wet) Pre-op NOM NOM 1974 NDM NDM 1975 NDM NDM 1976 NDM NOM 1977 NOM NOM 1978 NOM NDM 1979 NOM NDM 1980 NDM NDM 1981 NDM NOM 1982 NOM NOM 1983 101.8 NOM 1984 35.8 26.3 1985 46.7 21.1 1986 29 NDM 1987 69 15 1988 21.7 6.9 1989 NOM NOM 1990 NDM NDM 1991 NOM NDM 1992 NOM NOM 1993 NDM NOM 1994 NOM NOM 1995 NDM NOM 1996 NOM NOM 1997 NOM NOM 1998 NDM NDM 1999 NDM NDM 2000 NOM NOM 2001 NOM NOM 2002 NOM NOM 2003 NDM NOM 2004 NOM NOM 2005 NOM NDM 2006 NDM NOM 2007 NDM NDM 2008 NDM NDM 2009 NOM NOM 2010 NDM NOM 2011 NOM NOM 4-32
Table 4.7-2 (continued) A verage A nnua Ie s- 134 C oncen tra fIOn III . F*IS h Year Indicator Control (pCilk~-wet) (pCi/kg-wet) 2012 NOM NOM 4.8 Sediment Sediment was collected along the shoreline of the Altamaha River on May 21 and November 5, 2012, at the upstream control station (No . 170) and the downstream indicator station (No. 172). A gamma isotopic analysis was performed on each sample. Co-60 was detected in sediment samples near the plant from 1986, the year of the Chemobyl incident, through 2004. However, because Co-60 was detected in indicator station samples more often than in control station samples during the years 1986 through 2002, some contribution from plant effluents cannot be ruled out. Co-60 was not detected in 2012 and has not been detected in either control or indicator station samples since 2004. There is no RL or MOC assigned to Co-60 in sediment in OOCM Tables 4-2 and 4-3 (Tables 4-2 and 4-1 of this report). The MOC assigned by the EL for Co-60 in sediment is 70 pCi/kg-dry. The historical trending of the average annual detectable Co-60 concentration in sediment is provided in Figure 4.8-1 and Table 4.8-1 . 4-33
Figure 4.8-1 Average Annual Co-GO Concentration in Sediment 250 I 200 ~ "9 Cl ~ Ga. 150 c: .2
,\
-... 100 iQ c: Q) (J I \
~
V1 \ c:
\ II o
U 50 I 1\ I
""
I~! ~ I o f/ ~ '/ 1\/ ~ Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
--+-Indicator - t t - Control - - MOC 4-34
Table 4.8-1
. Sed*Imen t A verage A nnua Ie 0- 60 Concen t ra f IOn 10 Year Indicator Control (pCilkg-dry) (pCi/kg-dry)
Pre-op NOM NOM 1974 NDM NOM 1975 NOM NOM 1976 NOM NOM 1977 NOM NOM 1978 NOM NOM 1979 NOM NOM 1980 NDM NDM 1981 NOM NOM 1982 NOM NOM 1983 NDM NOM 1984 NOM NOM 1985 NOM NOM 1986 108 33 1987 NOM NOM 1988 67.8 NOM 1989 NDM 31 1990 33 19 1991 123.6 NOM 1992 81.4 NOM 1993 70.7 NOM 1994 218 NOM 1995 NOM NDM 1996 118.5 NOM 1997 NOM NOM 1998 79.4 NOM 1999 107.7 NOM 2000 70.0 NOM 2001 58.1 NOM 2002 NOM NOM 2003 NOM 31.5 2004 NOM NOM 2005 NOM NOM 2006 NOM NOM 2007 NOM NOM 2008 NOM NOM 2009 NOM NOM 2010 NOM NOM 2011 NOM NOM 4-35
Table 4.8-1 (continued) A verage A nnuaIe 0- 60 ConcentratlOn In . Se d*Iment Year Indicator Control (pCilk£-dry) (pCi/k£-dry) 2012 NOM NOM In 2012, Cs-137 was detected in both indicator and control station sediment samples from the spring collection. No radionuclides were detected in either the control or indicator station samples from the fall collection. Cs-137 has been found in over 95% of all of the sediment samples collected back through preoperation, and is generally attributed to atmospheric nuclear weapons tests or to the Chernobyl incident. As shown in Table 3-1, the concentration of Cs-137 detected in the indicator station sample from the spring collection was 30.3 pCilkg-dry and the concentration at the control station was 91.6 pCi/kg-dry. Because the value at the control station was higher than the value at the indicator station, it can be concluded that effluents from plant Hatch did not contribute to environmental concentrations. Due to the low number of samples, MOD was not able to be used to evaluate the data. The MDC for Cs-137 in sediment is 180 pCi/kg-dry. The historical trending of the average annual detectable Cs-137 concentration in sediment is provided in Figure 4.8-2 and Table 4 .8-2. 4-36
Figure 4.8-2 Average Annual Cs-137 Concentration in Sediment 1000 900 800 'E "'CI 700 Cl S l) 600 -c:: 0-500 \ 0
-....C1l 400 \ I / c:: Q) (J 300 I~ I ;1\ c:: 0 l) 1\ AV 1/ ~ 200 - 100
'\ \ I I
/~ ~ ~ I::::: Ll ,",
.....
~I ~ ~f ~v ./
"- "'1 1=<1 r-'
I I / I ~ ~h 0 . , Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
--Indicator ____ Control - - MDC 4-37
TabJe 4.8-2 A verage A nnua I C s- 137 c rIOn oncentra *III Sed*Iment Year Indicator Control (pCi/kg-dry) (pCi/kg-dry) Pre-op 170 270 1974 218 57 1975 330 615 1976 211 300 1977 364 200 1978 330 260 1979 NOM 310 1980 240 NOM 1981 590 110 1982 141 285 1983 384 365 1984 500 260 1985 76.5 269 1986 238 190 1987 59 39 1988 903 114 1989 56 62 1990 130.5 66 1991 43.1 54.5 1992 151 198.5 1993 113 115 1994 127 104 1995 52.3 80.6 1996 106 110 1997 186 137 1998 148.5 101.4 1999 92 111.8 2000 68.1 114.5 2001 68.7 69.6 2002 68.1 62.8 2003 57.3 106 2004 59.5 57.1 2005 57.2 30.3 2006 85.2 79.2 2007 82.1 71.6 2008 112.7 61.9 2009 74.9 60.5 2010 47.1 39.6 2011 40.2 61.2 4-38
Table 4.8-2 (continued) A vera2e A nnua Ie s- 137 C oncen t ra f IOn III . Se d'Iment Year Indicator Control (pCilkg-dry) (pCi/kg-dry) 2012 30.3 91.7 Other man-made nuclides, besides Co-60 and Cs-137, were occasionally found in past years. Their presence was generally attributed to the nuclear weapons tests or to the Chemobyl incident, although plant releases were not ruled out. Mn-54, Co-58, and Zn-65, which have relatively short half-lives, are most likely a result of plant releases and have been plotted in Figure 4.8-3 along with their MDCs. All the man-made nuclides detected in sediment except for Co-60 and Cs-137 have been listed in Table 4 .8-3. The Cs-134 MDC (150 pCi/kg-dry) is defined in ODCM Table 4-3 (Table 4-1 of this report) . The MDCs for Mn-54 (42 pCi/kg-dry) and Zn-65 (129 pCi/kg-dry) were determined by the EL since no values are provided in ODCM Table 4-3. Figure 4.8-3 Average Annual Indicator Station Concentrations of Select Nuclides in Sediment _500 1~~~4-~-+~+-~~~~~~4-~-+~+-~~4-~~~4-~-+~+-~
~
"0I ~400 l-~~4-~~~4-~-+~+4~~4-~~~+-~-+4-+-~~4-~~-r1-r1 U -o Q. C 300 ~-+~+-~~~+-~~4-~++~+-~~4-+-r+-r1-~-+-r+-~-+~+-r--l
III -~
~ 200 ~~4-~~++~~~~~+H~~fl\\~+~~~~~4-+4~4-~~+-r+~
C o
-I- -+-~~-t-++I-II-I--+-';-:~- I I +1-+-:.,....,+ -+;.+
U 100 1-4~~+-~-U~\~~\-Hu~ip~~~~++~-~~~~4-~~~4-~-+~+-~~~
\ 11\ ;: 1\ 1\ I\. \
o Po 75 77 79 81 83 85 87 89 91 93 95 97 99 03 05 07 09 11 Year
-Mn-54 -Zn-65 -..- Cs-134
-Mn-MDC --Zn-MDC - - Cs-MDC 4-39
Table 4.8-3 Sediment Nuclide Concentrations Other Than Co-60 & Cs-137 Nuclide YEAR Indicator Control (pCi/k~-dry) ~Ci/kZ-d~ Ce-141 1976 340 254 1977 141 Ce-144 Preop 720 1974 363 1975 342 389 1978 700 1981 1290 Co-58 1994 22.2 Cs-134 Preop 40 1981 280 1984 130 40 1986 132 1988 505 1990 31 Mn-54 1975 36.1 1986 28 26 1991 57.2 1996 77.7 Ru-l03 1974 81 1976 158 1977 195 1981 220 Zn-65 1986 175 1988 136 1991 250.5 1992 83 1993 39.9 1994 332 Zr-95 Preop 180 1974 138 1976 427 170 1977 349 294 1978 220 230 1981 860 280 4-40
4.9 Groundwater As nuclear plants began to undergo decommissioning in the late 1990s to early 2000s, instances of subsurface and/or groundwater contamination were identified. In addition, several operating facilities also identified groundwater contamination resulting from spills and leaks or equipment failure. In one instance, low levels of licensed material were detected in a private well located on property adjacent to a nuclear power plant. In 2006, NEI (Nuclear Energy Institute) fonned a task force to address monitoring onsite groundwater for radionuclides at nuclear facilities. A Groundwater Protection Initiative was developed which was adopted by all u.s. commercial operating nuclear plants. The NRC also formed a task force to study the groundwater issues and released Infonnation Notice 2006-13, Ground-water Contamination due to Undetected Leakage of Radioactive Water, which summarized its review of radioactive contamination of ground water at mUltiple facilities as a result of undetected leakage from structures, systems, and components that contain or transport radioactive fluids . Licensees were instructed to review the information for applicability and to consider appropriate actions to avoid similar problems. The NEI task force felt it was prudent for the industry to update site hydrology infonnation and to develop radiological groundwater monitoring plans at each site. These groundwater protection plans would ensure that underground leaks and spills would be addressed promptly. Additionally, the task force recommended developing a communications protocol to report radioactive leaks or spills that entered groundwater (or might eventually enter groundwater) to the NRC and State and Local government officials as needed. NEI-07-07, Industry Groundwater Protection Final Guidance Document, was developed by the task force to document the guidelines recommended for the industry. To ensure compliance with NEI-07-07, Southern Nuclear developed the Nuclear Management Procedure, Radiological Groundwater Protection Program. The procedure contains detailed site-specific monitoring plans, program technical bases, and communications protocol (to ensure that radioactive leaks and spills are addressed and communicated appropriately). The guidance in this procedure is used to infonnally update both the NRC and the State of Georgia regarding the changes in Hatch's groundwater tritium concentrations. In an effort to prevent future leaks of radioactive material to groundwater, SNC plants have established robust buried piping and tanks inspection programs. Plant Hatch has monitored onsite groundwater since preoperation. Initially piezometers, which were installed prior to plant construction, were used to monitor groundwater. In the late 1970s to the early 1980s timeframe, a hydrological engineering consultant was hired to evaluate several areas where leaks had occurred and tritium had been detected in onsite wells. The consultant recommended drilling additional monitoring wells to study the groundwater movement, to determine the source of the leaks, and to track the tritium concentrations in groundwater. The monitoring program continued over the years and most of the data generated in the late 1970s through the mid 1980s, was reported to the NRC. 4-41
The reporting frequency decreased over time for several reasons - the areas where the groundwater showed tritium were all onsite and the movement of groundwater was extremely slow and in a direction (towards the river) that was not expected to impact the public. During that period of time, there was no regulatory requirement to provide the reports. In 2006 as the nuclear industry was moving towards establishing groundwater protection programs, Plant Hatch hired a hydrological engineering consultant to re-evaluate the groundwater study which had been done previously. The key purpose of the new study was to evaluate the adequacy of the current monitoring program and to diagram the existing groundwater tritium plume to ensure that the plume had not migrated offsite. The consultant concluded that tritium was not leaving the site through the groundwater. The consultant recommended installing additional monitoring wells to better characterize the groundwater plume in areas of the site where there were no existing wells. During the course of Plant Hatch's groundwater evaluation in 2006, some leaks were discovered which explained why the levels of tritium around CST -1 (Unit 1 Condensate Storage Tank) were not decreasing. Underground piping which carried radioactive liquids was evaluated over the plant site and replaced in some areas around CST -1. Both CST tank/pump moats (Unit I and Unit 2) were coated and sealed to ensure that the moats would not leak in the event of transfer pump or tank leaks. In 2006, Plant Hatch's groundwater monitoring program included over 50 location points which were sampled on weekly, monthly, quarterly, or annual frequencies. Included in these sample points were the onsite drinking water wells (which did not contain detectable amounts of radioactivity above background). Surface drains or outfalls were also included as sample points. Tritium was detected in two of the outfalls which discharged to the river. These outfalls were initially added to the Hatch ODCM as radiological effluent release points. Pennitted release point Y22N008A (by design) discharges groundwater from the site subsurface drainage system which includes the tritiated groundwater around the CST-l. The other release point, Y22N003A, discharges runoff from the roof drains. The source of tritium in this outfall has been detennined to be from rain washout of the gaseous plant effluents and is no longer a permitted release point. Plant Hatch sampled rainfall during two rain events in 2006 and found tritium levels as high as 4.58E5 pCi/1 on the reactor building roof. Two other outfalls, Y22N024A and Y22N025A, which discharge into the onsite swamp show sporadic levels of tritium. The source of tritium in these outfalls is also believed to be from rain washout. In 2007, Hatch continued to aggressively monitor the groundwater tritium plume especially in two areas of higher activity around CST -1 and CST -2. The amount of seasonal rainfall during 2007 seems to have had some correlation with the tritium concentrations in the T-12 well near CST-I. During early spring and late fall rainy seasons, the concentrations of tritium were at their highest levels, whereas, during the summer and early fall drought season the tritium concentrations decreased significantly. This is indicative of water table level fluctuations . However, this same seasonal affect was not observed in the newer NWIO monitoring well installed in 2006 near CST -2. The tritium concentration in NW 10 increased from February 2007 through September 2007 by a factor of2.5. 4-42
Events which could have contributed to the increase were a CST -2 transfer pump leak (in November 2006) which led to an accumulation of a couple feet of CST-2 water in the pump moat. Although the moat had been sealed earlier in 2006, there was a possibility that some of the contaminated water seeped through the concrete moat and gradually seeped through the ground to NW 1O. In addition, there was a deep hole dug (in January 2007) near the CST -2 (and NW 10) to replace some CST-2 piping. The hole may have altered groundwater flow toward NWI0 from the CST -1 groundwater plume and resulted in higher concentrations of tritium being drawn to NWlO. In 2008, Hatch made further enhancements to the groundwater tritium monitoring program. Three additional shallow wells and three additional deep wells were installed ("R" series wells). One of the deep wells was a replacement well for the deep well N7 A. The integrity of N7 A was questioned due to the high level of tritium (-211,000 pC ill) seen in this well which should have been protected from contamination by a confining layer. The well was retired and a new well (R-3) was placed in the same vicinity. The newer well showed much smaller amounts of tritium activity (average of 1324 pCi/1 in 2012) . In addition, several other groups within Southern Company are now utilized to conduct an improved sampling program and to provide additional expertise in characterizing groundwater quality and flow. The sampling frequency for radiological groundwater monitoring was officially changed to quarterly starting in second quarter of 2008 with SCS Civil Field Services perfonning the sampling and Georgia Power Environmental Laboratory continuing to analyze the samples. Southern Nuclear Corporate Engineering and Hatch Site Engineering have developed a Buried Piping and Tanks Inspection Program. This program should help to prevent releases of radioactive material to groundwater. Underground piping and components are risked ranked using detailed procedures and EPRI's software, BPWorks, to ensure vulnerable areas are identified and repaired or replaced before problems occur. In May of 2009, there was an increase in tritium concentration in well T-3 (located near the U-l Turbine Building) from approximately 2600 pCi/1 to approximately 37,000 pCi/1. Neighboring well N9B (not part of the formal GW sampling program) also showed an approximate lOX increase - going from 1300 pCi/1 to over 10K pCi/l. Investigation found no process leaks and the non-rad constituents continued to match groundwater. The increase was attributed to migration of the plume. Increased rainfall and the fact that the wells are located near the subsurface drain could likely have facilitated the pathway of the plume towards the T3 well. A courtesy notification was made to the State of Georgia Dept. of Natural Resources, and a 10CFR50.72 formal report was made to the NRC - although only courtesy notifications were required per procedure. In 2012, the quarterly tritium concentrations in T3 ranged from 1840 pCi/1 to 6320 pC il l during the year (average of 3282 pCi/1 in 2012) but remained below the established Administative Control Limit (ACL) of 37,000 pCi/1. Administrative Control Limits (ACL) were established near the end of 20 I 0 for the surficial and deep aquifers and for specific wells based on the presence of legacy tritium, the previous well results, and total measurement uncertainty. There are no reporting requirements associated with exceeding an ACL but additional actions would be taken to verify no new sources of tritium if an ACL was exceeded. The ACL for T-12 is 900,000 pCi/1 and the average for T-12 in 2012 was 99,000 pCi/1 of tritium with a range in values of 50,500 to 212,000 pCi/1 (157,775 pCi/ 1 was the 4-43
2011 average in T-12). For NW-I0, the ACL is 160,000 pCi/\ and the average tritium concentration in 2012 was 15,075 pCi/\ with a range of 12,000 to 17,600 pCi/1 (22,225 pCi/1 was the average in 20 II). These two wells, in areas of legacy contamination, continue to trend downward. The ACL for T 10 (25,000 pCi/l) was exceeded on 9/21 Ill. The result, which was detennined on 9/28111, was 4.61E6 pCi/1 and the previous sample had been only 5000 pCi/1. Additional samples were taken to verify the high tritium level and hard to detect radionuclides, Sr-89/90 and Fe-55, were analyzed for but were detennined to be at background levels. No gamma emitters were detected. Voluntary communications with state/local stakeholders were perfonned on 9/29111. A fonnal 10CFR50. 72b notification to NRC was also made. A response team was assembled (with 24 hour coverage) to identify the source of the tritium. Buried piping that supplied water from the CST -I to the liquid radwaste processing system was identified as the source, and the section of piping was evacuated of water and abandoned in place. A design change was completed to replace it with aboveground piping. No drinking water sources were impacted by the leak, and the tritium was contained within the shallow perched aquifer located in the immediate vicinity of CST-I. Two new wells, NU-1 and NU-2, were drilled in the area near the leak to enhance monitoring and to facilitate remediation activities (pumping out wells to remove the contamination). By 4th quarter sampling (12/06111), the tritium in T-1O had decreased to 2.3E6 pCi/1. Sample results for this location in 2012 have averaged 61,766 pCi/1. No tritium activity above background has been detected in the Deep Wells 2 and 3 which are used for drinking water at the plant. The plant staff continues to sample and monitor strategically located wells on a more frequent basis than quarterly to ensure that radiological leaks have not occurred. In addition, outfalls, pull boxes, manholes, and the sewage treatment plant effluent are sampled by the plant staff on a periodic basis. The latest groundwater tritium plume map (generated from the 2012 SCS 4th quarter sampling data) is shown as Figure 4.9-1 on the last page of this section. It is a representation of the current groundwater conditions at Plant Hatch. The two wells of interest around the CSTs (T -12 and NW -10) continued to show an overall decreasing trend as discussed above. The subsurface drain system and rainfall continue to influence groundwater movement around the site and contribute to the wide range of tritium values seen in the groundwater monitoring wells. In 2012, there was one leak detected by routine onsite groundwater sampling. On December 19, 2012 the samples from the T II and T 12 wells near CST-I had tritium concentrations of 4,800,000 pCi/1 and 5,700,000 pCi/l, respectively, approximately 100 times the nonnal values. Additional samples were taken from the same points and analysis confinned that tritium was present at elevated levels in both samples and that tritium had not migrated out of the general area of initial discovery. The investigation showed that the tritium is confined to a small area on the plant site in the vicinity of CST-I and there is no significant potential for off-site impact. These samples and results took place after the data indicated in the plume map mentioned above. Further investigation revealed the source of the tritium was a transfer pipe associated with CST-I. The leak has been repaired and additional monitoring has been initiated. The December 31,2012 tritium concentration at TIl was 7,360,000 pCi/l, and the concentration at Tl2 was 6,480,000 pCi/1. 4-44
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5.0 INTERLABORATORY COMPARISON PROGRAM In accordance with aDCM 4.1.3, the EL participates in an ICP that satisfies the requirements of Regulatory Guide 4.15, Revision I, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment", February 1979. The guide indicates the ICP is to be conducted with the Environmental Protection Agency (EPA) Environmental Radioactivity Laboratory Intercomparison Studies (Cross-check) Program or an equivalent program, and 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 Quality Assurance (QA) program and the capability to prepare Quality Control (QC) 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 and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses of water samples. 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 as reported by Analytics. The uncertainty of the reported average is the propagated error of the values in the reported average by the EL. The precision of each result is measured by the coefficient of variation, which is defined as the standard deviation of the reported result 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-Sl and Fe-59. For Cr-51 and Fe-59, an investigation is undertaken when the coefficient of variation exceeds the values shown as follows: Nuclide Concentration
- Total Sample Activity Percent Coefficient (pCi) of Variation Cr-5l <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 umts are pCI/filter. For all other medIa, concentration units are pCi/liter (pCi/I).
5-\
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 and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses 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. The Environmental Radiochemistry laboratory participates in a performance evaluation (PE) sample program provided by Analytics Inc. The PE samples are received and analyzed routinely with environmental and effluent samples. The laboratory analyzed 9 samples for 35 parameters in 2012. The 2012 analyses included tritium, gross beta and gamma emitting radio-nuclides in different matrices. The attached results indicate 3 analyses (Ce-141, Cr-51, and Fe-59) were outside the acceptance limits for accuracy. These isotopes were in the Gamma in Air Filter matrix. After the results were received, the sample was recounted but two of the isotopes had decayed off. The remaining isotopes were within acceptable limits for accuracy. A Gamma in Air Filter PE sample will be analyzed in 2Q 2013 to complete an investigation . 5-2
TABLE 5-1 (SHEET lof3) INTERLABORATORY COMPARISON PROGRAM RESULTS 1-131 ANALYSIS OF AN AIR CARTRIDGE (pCi/cartridge) Analysis or I Date Reported Known I Standard I Uncertainty I Percent Coef I Normalized Radionuclide Prepared I Average I Value Deviation EL Analytics (3S) of Variation Deviation 1-131 I 07114112 I 100.00 I 97.20 I 6.45 I 1.62 I 7.22 I 0.44 GAMMA ISOTOPIC ANALYSIS OF AN AIR FILTER (pCi/filter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Ce-141 09/13112 113.00 153.00 6.66 2.56 7.72 0.02 Co-58 09113112 77.50 94.10 3.85 1.57 7.41 -2.90 Co-60 09113112 117.00 142.00 4.14 2.38 5.69 -3.82 Cr-51 09113112 184.00 232.00 18.3 3.88 14.27 -1.84 Cs-134 09/13112 83.40 101.00 2.73 1.69 5.39 -3.92 , Vl w Cs-137 09113112 130.00 163.00 4.83 2.73 5.96 -4.22 Fe-59 09113112 111.00 142.00 7.18 2.38 8.76 -3.14 Mn-54 09113112 150.00 183.00 12.5 3.06 9.49 -2.29 Zn-65 09113112 154.00 180.00 13.4 3.01 10.16 -2.29 GROSS BETA ANALYSIS OF AN AIR FILTER (pCi/filter) Analysis or Date I Reported I Known I Standard I Uncertainty I Percent Coef I Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 09113112 I 93.00 I 84.10 I 1.2 I 1.40 I 4.21 I 2.27
TABLE 5-1 (SHEET 2 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF A MILK SAMPLE (pCiJliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Ce-141 7/ 14112 76.60 82.20 2.85 1.37 8.62 -0.84 Co-58 7/ 14112 90.20 99.70 2.85 1.66 7.33 -0.32 Co-60 7114/ 12 350.00 355.00 10.7 5.93 4.23 -0.35 Cr-51 7114112 433.00 402.00 21.1 6.71 8.51 0.85 Cs-134 7/14112 180.00 174.00 6.26 2.91 4.68 0.69 Cs-137 7114112 216.00 212.00 9.26 3.54 5.63 0.34 Fe-59 7/ 14/ 12 126.00 128.00 6.53 2.13 8.21 -0.17 1-131 7114/ 12 102.00 99.70 7 1.66 8.78 0.30 Mn-54 7114112 134.00 132.00 3.79 2.21 5.69 0.24 Zn-65 7114112 208.00 199.00 8.17 3.33 7.09 0.59 GROSS BETA ANALYSIS OF WATER SAMPLE (pCiJliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 03115112 263.00 297 .00 18.93 4.96 1.10 -11. 99 07114112 166.00 148.00 10.28 2.47 10.85 0.98 GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES (pCiJliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL AnalyHcs (3S) of Variation Deviation Ce-141 03/15112 190.00 184.00 5.65 3.07 5.25 0.64 Co-58 03115112 92.50 93.40 5.19 1.56 8.20 -0.11 Co-60 03115112 208.00 197.00 5.68 3.29 4.59 1.16
TABLE 5-1 (SHEET 3 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES CONT. (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Cr-51 03/15112 362.00 309.00 46.7 5.16 15.63 0.94 Cs-134 03115112 108.00 106.00 1.75 1.77 4.41 0.49 Cs-137 03115112 124.00 113.00 3.67 1.88 6.09 1.50 Fe-59 03115/12 119.00 119.00 6.14 1.99 8.23 0.02 1-131 03/15112 104.00 93.80 4.15 1.57 6.68 1.44 Mn-54 03/15112 149.00 138.00 2.24 2.31 5.13 1.38 Zn-65 03115112 245.00 235.00 5.41 3.93 5.98 0.67 TRITIUM ANALYSIS OF WATER SAMPLES (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized
,
IJI IJI Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation H-3 03115112 4160 4470 102.54 74.70 4.43 -1.70 07114112 4580 4970 92.36 83.00 4.34 -1.98
6.0 CONCLUSION
S This report confirms the licensee's conformance with the requirements of Chapter 4 of the ODCM. It provides a summary and discussion of the results of the laboratory analyses for each type of sample. In 2012, there were no instances where the indicator station results were statistically discernible from the control station results. 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. The REMP trends over the course of time from preoperation to the present are decreasing or have remained fairly constant. This supports the conclusion that there is no adverse radiological impact on the environment or to the public as a result of the operation of Hatch Nuclear Plant. 6-1
7.0 ERRATA The following pages are corrections to the Edwin I. Hatch Nuclear Plant Annual Radiological Environmental Operating Report for 20 II . The corrections are a result of the discovery, by Georgia Power Company Environmental Laboratory staff in 2012, of a small positive bias in the 2011 results of OSL environmental dosimeter readings. The method used during 20 II was acceptable at the time but EL dosimetry personnel studied the source of the bias and determined it was based on higher residual dose on the OSL badges as compared to the past Panasonic system . New processing methods are now in place and included in processing procedures. All 2012 environmental OSL processing and reports have included the new methods to remove this small positive bias. The correction has been applied to the 20 II OSL dosimeter results and the corrected data are described in the following pages. Additional minor changes resulting from OSL dosimeter data review included corrections to a mis-identified OSL dosimeter on the sample deviation table 4-3. During the third quarter, the OSL dosimeter was missing at Station #105, not Station
# 115.
The text description of historical trending of direct radiation monitoring at areas of special interest has been changed to include "since 1986". 7-1
TABLE 3-1 (SHEET 1 of 4) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Edwin I. Hatch Nuclear Plant, Docket Nos. 50-321 and 50-366 Appling County, Georgia Medium or Type and Total Minimum Indicator Location with the Highest I, Other Control Pathway Number of Detectable Locations Annual Mean Stations(g) Locations Sampled Analyses Concentration Mean (b), I Mean (b), Mean (b), (Unit of Performed (MDC) (a) Range Name Distance Mean (b), Range Range Measurement) (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 23.5 No. 309 26.1 NA 25.1 Particulates 312 6.4-43.5 Baxley 9.4-57.1 5.2-57.1 (fCi/m3) (208/208) Substation (52/52) (1041104) _.*.. __._._-_._-- _.........._-_....... 10 miles S --_ _- ----- -----._-"
..
Gamma Isotopic 24 Cs-134 50 NDM (c) NDM NDM Cs-137 60 NDM NDM NDM Airborne 1-131* 70 56.7 No. 304 80.4 NA 66.8 Radioiodine 312 16.9-101.2 State Prison 58.2-102.6 19.9-106.9 (fCi/m3) (111208) 11.2 miles (2/52) (51104) -....I W I I CORRECTION ~312 ENE N - FOR 2010 Direct Radiation Gamma Dose NA(d) 13.7 No. 115 18.2 13.1 13 .1 (mRI91 days) 146 9.7-19.0 Inner Ring 17.4-18.9 9.0-19.1 10.3-14.5 (63/63) 1.1 miles NW (4/4) (71171) (12/12) Milk Gamma Isotopic (pCi/l) 25 Cs-134 15 NA NDM NA NDM Cs-137 18 NA NDM NDM Ba-140 60 NA NDM NDM La-140 15 r*-----***-----****-*--*-*-- --***** NA NDM NDM ---------" 1-131 NA NDM NDM 25
TABLE 4-3 DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLES 01/03 / 11-01 / 12/11 Air and Milk samples Sample collection Ice storm delayed sample collection Samples were collected delayed from 0 II I0111 when conditions were safe to 01112/11 for travel 03121111-03128111 AC/AF N on-representati ve Lost 10.6 hours of sample time due to Repaired by Georgia Power Station #309 sample of airborne electrical issue with air station caused Electrical Engineering 10.0 miles S particulates by lightning strike Dept. 15 1 Quarter OSL Dosimeters Non-representati ve Badges were missing at collection time Replaced OSL dosimeters Station #214 direct radiation data at beginning of quarter 5.4 miles WNW 04/25111-05/02111 AC/AF Non-representative Lost 4 hours of sample time due to Power restored after Station # 112 sample of airborne maintenance activities on power supply maintenance perforn1ed 1.0 mile WSW particulates yd Quarter OSL Dosimeters Non-representati ve Badges were lost in transit to lab Replaced OSL dosimeters Station # 105 direct radiation data at beginning of quarter 3.7 miles E 3fd Quarter OSL Dosimeters N on-representati ve Water in holder packages found at Replaced OSL dosimeters Station #213 direct radiation data collection time at beginning of quarter 4.3 miles W 11107/ 11-12/05/11 River Water Non-representative Problem with ISCO autosampler Collected a grab sample Station 172 monthly river water 3 miles Downstream composite NOTE: Hatch CR 451553 documents the REMP deviations for 2011
4.3 Direct Radiation In 2011, direct (external) radiation was measured with Landauer InLight optically stimulated luminescent (OSL) dosimeters which replaced the Panasonic thermoluminescent dosimeters (TLOs). The Panasonic system was retired at the end of 2010 due to the inability to keep the aging badge readers operating reliably. Similar to the TLO protocol of the past, two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges. Two direct radiation stations are established in each of the 16 compass sectors around the plant to form 2 concentric rings, as seen in Figures 2-1 and 2-2. The two ring configuration of stations was established in 1980, in accordance with NRC Branch Technical Position "An Acceptable Radiological Environmental Monitoring Program", Revision 1, 1979. With the exception of the East sector, the inner ring stations (Nos. 101 through 116) are located near the site boundary and the outer ring stations (Nos. 201 through 216) are located at distances of 4 to 5 miles from the plant. The stations in the East sector are a few miles farther out than the other stations in their respective rings due to large swamps making normal access extremely difficult. The 16 stations forming the inner ring are designated as the indicator stations. The 3 control stations (Nos. 304, 309 and 416) are located 10 miles or more from the plant. Stations 064 and 301 monitor special interest areas. Station 064 is located at the onsite roadside park, while Station 301 is located near the Toombs Central School. Station 210, in the outer ring, is located near the Altamaha School (the only other nearby school). As provided in Table 3-1, the average quarterly exposure measured at the indicator stations (inner ring) during 2011 was 13.7 mR. At the control stations, the average quarterly exposure was 13 . 1 mR. This difference (0.6 mR) is not statistically discernible since it is less than the MOD of 1.6 mR. The quarterly exposures acquired at the outer ring stations during 2011 ranged from 9.0 to 19.1 mR, with an average of 13.1 mR. The average for the outer ring stations was the same as the average for the control stations. Since the results for the outer ring stations and the control stations differ by less than the MDO of 1.6 mR, there is no discernible difference between outer ring and control station resul ts for 2011. 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 TLOs from Teledyne to Panasonic. It should be noted however that the differences between indicator and control and outer ring values did not change. 4-12 7-4
During 2010, OSL badges were co-located on station with the TLD badges. In 2011 , only the OSL badges were placed at each station. Following the change to only OSL badges, the differences between indicator, control, and community locations has been consistent with previous years. An increase noted in 2010 n reflects issues (especially during 2 Qtr) with the aging Panasonic TLD reader. The close agreement between the station groups supports the position that the plant is not contributing significantly to direct radiation in the environment. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 30 25 v V _ 20 et:: .s
~ 15
- l I/)
~
v..., . . .
}\
H
~ ~.~
)-l
\ V
,
I' oQ. i~1, :r-- ~ .~ -;::::: ~ r-'/.'
- I
~ 10
~ ....,
~ I i::; k:\ ""'" L 5
o Po 75 77 79 81 83 85 87 89 91 93 95 97 99 01 03 05 07 09 11 Year
--Indicator _ Control - - Outer Ring I 4-13 7-5
Table 4.3-1 Average Quarterly Exposure from Direct Radiation Year Indicator (mR) Control (mR) Outer rung (mR) Pre-op 2203 23 NA 1974 2302 2506 NA 1975 1000 1005 NA 1976 8018 609 NA 1977 7031 6052 NA 1978 6067 6001 NA 1979 5016 6.77 NA 1980 4.44 5.04 4.42 1981 509 5.7 5.7 1982 1203 12 11.3 1983 11.4 11.3 1006 1984 1303 12.9 1l.9 1985 1407 14.7 13.7 1986 15 14 14.5 1987 14.9 14.6 15.3 1988 15.0 14.7 1502 1989 16.4 1800 16.5 1990 1409 1309 1407 1991 1501 1307 1506 1992 11.9 1009 1203 1993 11.6 1007 11.5 1994 11 1007 11.2 1995 11.5 1008 11.3 1996 11.6 11.3 11.6 1997 1203 11.8 1203 1998 1201 1203 12.3 1999 1208 13.2 13 00 2000 1306 13.3 13 .3 2001 1200 12.1 11.8 2002 11. 7 11.7 11.5 2003 11.4 11.4 11.4 2004 1202 12.4 12.2 2005 1201 1205 12.0 2006 12.4 11.9 1108 2007 1208 12.5 12.6 2008 1300 1203 12.4 2009 12.4 1202 12 02 2010 1508 1506 1600 2011 1307 13.1 13.1 4-14 7-6
The historical trending of the average quarterly exposures at the special interest areas since 1986 is 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 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 25 20 ~ 15
~ ~ ..... ......
-E Q) 1/1 1\ ~
~v f...---l
.- ...- ~
~~ ~ V '\
o Cl 10 5 o 86 87 88 89 90 91 92 93 94 95 96 97 98 99 0 1 02 3 4 5 6 7 8 9 10 11 Year
-+- Roadside Park (Sta 064) ___ Toombs Central School (Sta 301) 4-15 7-7
Table 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas Period Station 064 Station 301 (mR) (mR) 1986 14.6 15.1 1987 14.2 15.0 1988 14.9 15.3 1989 16. 1 16.6 1990 15.1 14.4 1991 14.4 15.2 1992 11.1 11.5 1993 11.2 10.8 1994 10.4 10.7 1995 11.0 10.5 1996 11.7 11.0 1997 12.6 11.4 1998 12.4 11.8 1999 12.5 12.4 2000 13.3 12.6 2001 11.8 11.3 2002 11.4 11.4 2003 11.2 1l.1 2004 11.9 12.3 2005 11.8 12.4 2006 1 1.9 11.6 2007 11. 9 12.1 2008 12.3 12.2 2009 12.1 12.1 2010 15.7 15 .5 2011 12.7 13.1 Table 4-3 lists the REMP program deviations that occurred in 2011. There were three deviations involving OSL dosimeters. At first quarter collection, badges were missing at Station #214. At third quarter collection, badges at Station # 105 were lost during transit after collection, and badges at Station #213 had water in the holding bag but the data passed Chauvenet's Criterion and was retained in the data set. 4-16 7-8
The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self-imposed limit of 3.5. Previously with TLDs, this limit had been 1.4. However, the OSL readings varied more (between the two elements) than the TLD readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication ISO, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 20 I O. The limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater or equal to 3.5 are excluded from the data set since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 20 II, the following OSL results were excluded from the data set because their standard deviations were greater than or equal to 3.5: First Quarter HI03B,H204B,H205B,H30IB Second Quarter HI02A, HI09B, H203B, H212B Third Quarter H210B Fourth Quarter None If one badge at a station exhibited a standard deviation greater than or equal to 3.5, then the reading of the companion badge at each location would be used to determine the quarterly exposure. 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 found for the high standard deviations. A major advantage of the OSL badge is that it can be read multiple times. A new practice was employed in 2011 to re-read any environmental badges that yielded a standard deviation 2: 3.5. The readings with the lower standard deviation would be reported. 4-17 7-9
Edwin I. Hatch Nuclear Plant Joseph M. Farley Nuclear Plant Vogtle Electric Generating Plant Annual Radiological Environmental Operating Reports for 2012 Enclosure 2 Farley Annual Radiological Environmental Operating Reports for 2012
JOSEPH M. FARLEY NUCLEAR PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 SOUTHERN COMPANY Energy to Serve Your World'"
JOSEPH M. FARLEY NUCLEAR PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 April 26,2012 FINAL "Chern Tim Meents Ian Lake Tim.Meents@chemstaff.com lan.Lake@chemstaff.com 815-600-9247 815-600-2067 Dennis Oltmans DOltmans@chemstaff.com 717-575-3481
T ABLE OF CONTENTS Section and/or Title Subsection Page List of Figures 11
...
List of Tables 111 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 4-5 4.2 Airborne 4-6 4.3 Direct Radiation 4-13 4.4 Milk 4-17 4.5 Forage 4-21 4.6 Ground Water 4-27 4.7 Surface Water 4-32 4.8 Fish 4-35 4.9 Sediment 4-41 5.0 Interlaboratory Comparison Program 5-1 6.0 Conclusions 6-1 7.0 Errata 7-1
LIST OF FIGURES Fi~ure Number Title Pa~e Figure 2-1 REMP Stations Near the Plant Perimeter 2-10 Figure 2-2 REMP Stations 2 to 5 Miles from the Plant 2-11 Figure 2-3 REMP Stations Beyond 5 Miles from the Plant 2-12 Figure 2-4 Onsite Ground Water Monitoring Locations 2-13 Fi~re 4.2-1 Average Weekly Gross Beta Air Concentration 4-7 Figure 4.2-2 A verage Annual Cs-13 7 Concentration in Air 4-10 Figure 4.3-1 A verage Quarterly Exposure from Direct Radiation 4-14 Figure 4.4-1 A verage Annual Cs-137 Concentration in Milk 4-17 Figl,lre 4.4-2 Average Annual 1-131 Concentration in Milk 4-19 Figure 4.5-1 Avera~e Annual Cs-137 Concentration in Fora~e 4-22 Figure 4.5-2 Average Annual 1-131 Concentration in Forage 4-24 Figure 4.6-1 Average Annual H-3 Concentration in Offsite Ground 4-28 Water Figure 4.6-2 H-3 Concentration in Onsite Ground Water Well R-3 4-31 Figure 4.7-1 Average Annual B-3 Concentration in Surface Water 4-33 Figure 4.8-1 Average Annual Cs-137 Concentration in Bottom FeedinKFish 4-36 Figure 4.8-2 A verage Annual Cs-13 7 Concentration in Game Fish 4-38 Figure 4.9-1 Average Annual Cs-134 Concentration in Sediment 4-42 Figure 4.9-2 Average Annual Cs-13 7 Concentration in Sediment 4-43 11
LIST OF TABLES Table Number Title Page Table 2-1 Summary Description of Radiological Environmental Monitoring Program 2-2 Table 2-2 Onsite Groundwater Monitoring Locations 2-9 Table 3-1 Radiological Environmental Monitoring Program Annual Summary 3-2 Table 4-1 Minimum Detectable Concentrations (MDC) 4-1 Table 4-2 ReportingLevels CRL) 4-2 Table 4-3 Deviations from Radiological Environmental Monitoring Program 4-4 Table 4.1-1 Land Use Census Results 4-5 Table 4.2-1 Average Weekly Gross Beta Air Concentration 4-8 Table 4.2-2 A verage Annual Cs-13 7 Concentration in Air 4-11 Table 4.3-1 A verage Quarterly Exposure from Direct Radiation 4-15 Table 4.4-1 A verage Annual Cs-13 7 Concentration in Milk 4-18 Table 4.4-2 Average Annual 1-131 Concentration in Milk 4-20 Table 4.5-1 Average Annual Cs-13 7 Concentration in Forage 4-23 Table 4.5-2 A verage Annual 1-131 Concentration in Forage 4-25 Table 4.6-1 Average Annual H-3 Concentration in Ground Water 4-29 Table 4.7-1 Average Annual H-3 Concentration in Surface Water 4-34 Table 4.8-1 Average Annual Cs-13 7 Concentration in Bottom Feeding Fish 4-37 Table 4.8-2 Average Annual Cs-137 Concentration in Game Fish 4-39 Table 4.9 Sediment Nuclide Concentrations 4-41 Table 5-1 Interlaboratory Comparison Program Results 5-4 III
LIST OF ACRONYMS Acronyms presented in alphabetical order Acronym Definition APCo Alabama Power Com~any ASTM American Society for Testing and Materials CL Confidence Level EL Georgia Power Company Environmental Laboratory EPA Environmental Protection Agency FNP Joseph M. Farley Nuclear Plant ICP Interlaboratory Comparison Program MDC Minimum Detectable Concentration MDD Minimum Detectable Difference MWe MegaWatts Electric NA Not Applicable NDM No Detectable Measurement(s) NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual OSL Optically Stimulated Luminescence Po Preoperation PWR Pressurized Water Reactor REMP Radiological Environmental Monitoring Program RL Reporting Level RM River Mile SNOC Southern Nuclear Operating Company TLD Thennoluminescent Dosimeter TS Technical Specification IV
1.0 INTRODUCTION
The Radiological Environmental Monitoring Program (REMP) for 2012 was conducted in accordance with Chapter 4 of the Offsite Dose Calculation Manual (ODCM). The REMP activities for 2012 are reported herein in accordance with Technical Specification (TS) 5.6.2 and ODCM 7.1. The objectives of the REMP are to:
- 1) Detennine 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 Joseph M. Farley Nuclear Plant (FNP).
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), areas of higher population (community 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. FNP is owned by Alabama Power Company (APCo) and operated by Southern Nuclear Operating Company (SNOC). It is located in Houston County, Alabama approximately fifteen miles east of Dothan, Alabama on the west bank of the Chattahoochee River. Unit 1, a Westinghouse Electric Corporation Pressurized Water Reactor (PWR) with a licensed core thennal power output of 2775 MegaWatts thennal (MWt), achieved initial criticality on August 9, 1977 and was declared "commercial" on December 1, 1977. Unit 2, also a 2775 MWt Westinghouse PWR, achieved initial criticality on May 8, 1981 and was declared "commercial" on July 30, 1981. The preoperational stage of the REMP began with initial sample collections in January of 1975. The transition from the preoperational to the operational stage of the REMP was marked by Unit 1 initial criticality. A description of the REMP is provided in Section 2 of this report. An annual summary of the results of the analyses of REMP samples is provided in Section 3. A discussion of the results, including assessments of any radiological impacts upon the environment and the results of the land use census are provided in Section 4. The results of the Interlaboratory Comparison Program (ICP) are provided in Section 5. Conclusions are provided in Section 6.
] -l
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 perfonned 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 describes the locations of the indicator, community and control stations as described in ODCM Table 4-4 and the identification of each sample according to station location and analysis type. The stations are also depicted on maps in Figures 2-1 through 2-4. The location of each REMP station for gaseous releases is described by its direction and distance from a point midway between the Unit 1 and Unit 2 plant vent stacks. The surrounding area is divided into 16 azimuthal sectors which are centered on the major compass points; each sector is numbered sequentially clockwise and oriented so that the centerline of sector 16 is due north. Each sampling station is identified by a four digit number. The first two digits indicate the sector number, and the last two digits indicate the distance from the origin to the nearest mile. For example, air monitoring station 0215 is located approximately 15 miles northeast of the origin. The locations for the sampling stations along the river are identified by the nearest River Mile (RM) which is the distance along the navigable portion of the Chattahoochee River upstream of the Jim Woodruff Dam near Chattahoochee, Florida. The approximate locations of the plant discharge and intake structures are at RM 43.5 and 43.8, respectively. The samples are collected by the plant's technical staff, except for fish and river sediment samples which are collected by APCo Environmental Field Services personnel. All laboratory analyses were perfonned by Georgia Power Company's Environmental Laboratory (EL) in Smyrna, Georgia. 2-1
TABLE 2-1 (SHEET 1 of 7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Identification Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types and Locations (sector-miles) AIRBORNE Particulates Continuous sampler operation with sample Particulate sampler: Analyze for gross collection weekly. beta radioactivity 2 24 hours following filter change. Perform gamma isotopic analysis on each sample when gross beta activity is > 10 times the yearly mean of control samples. Perform gamma isotopic analysis on composite ______~-~.----~------.---~-----------------------~mpk~y~c~~~gw~l~ _ Indicator Stations: River Intake Structure PI-OSOI (ESE-0.8) tv South Perimeter PI-0701 N (SSE-I .O) Plant Entrance PI-IIOl (WSW-0.9) North Perimeter PI-160l (N-0.8)
- _._-- -- --;--._ - --+- - - - - - --- -_..._--_._. _._-_._._ . _. . .----.- . -------.. - .-- -.---.----.-. - .---- ------ - -~- ------------ -----
Control Stations: Blakely GA (NE-IS) PB-02IS Neals Landing, FL PB-0718 (spare (SSE-18) station, not in service)
~~n, AL JW -18L+--P..:::..B_-I_2_1_8____ _ --+__.___._. __._. __. ____._._. ____. _. ____. ._. ____.____ . . . _. . . _ ____ _ ___ _.___ ___ _ _ ____ _ _.__ _ __ __
Community Stations: GA Pacific Paper Co. PC-0703 (SSE-3) Ashford, AL PC-II 08 (WSW-8) Columbia, AL (N-S) PC-l60S
TABLE 2-1 (SHEET 2 of7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Identification Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types I and Locations (sector-miles} I Iodine Continuous sampler operation with sample Radioiodine canister: Analyze each
- - - ---
__~gJ)_~g!~2~ _~~els)j'________ ___________ _ __ __________ sample for I-l31 weekly. 1D.-dicator Stations: River Intake Structure II-OS01 (ESE-O.8) South Perimeter II-070 1 (SSE-l.O) Plant Entrance II-I 101 (WSW-O.9) North Perimeter II-1601 r ili-M)
- --- . ----- - - - --- __ .*_ _ .....*.._ .* _*** * .___.******v.... _ _ *__ ***.*.._*_ _.... __ ._ _ ._ __. _ _ _.. _ .. __ ._ _ .*. __ _ _ .* __ __ ._ _.
--- -
Control Station: Blakely, GA (NE-1S) IB-021S Neals Landing, FL IB-0718 (spare station, (SSE-I8) not in service) Dothan, AL (W -18) IB-I218
**w _ ,--- ---... _--_..'._----_ __ _-------- .. .. _._-.--_ ..* -, , -..*..... _ _. __ __....__._--_._-_
..*. ._. .. -_._._-_._---
'Community Station: GA Pacific Paper Co. IC-0703 (SSE-3) DIRECT RADIATION Quarterly Gamma dose: Read each badge
- - ....-..__..-._--_... _._.__ ....... -.- -_._.._----_...._.-_.,-_._-_._-------------_ .. _----. quarterly -
Indicator Stai-ioi1S:- -- Plant Perimeter (NNE-O.9) RI-O 10 1 (NE-l.O) RI-0201 (ENE-O.9) RI-030I (E-O.8) RI-040I
.(ESE-O.8) RI-OS01
TABLE 2-1 (SHEET 3 of 7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types Identification and Locations (secto r-miles) (SE-I.l) Rl-060l (SSE-l .O) Rl-070l (S-l.O) Rl-080l (SSW-1.0) Rl-090l (SW-O.9) Rl-lOOl (WSW-O.9) Rl-llOl (W-O.8) Rl-120l (WNW-O.8) Rl-130l (NW-1.l) RI-140l (NNW-O .9) Rl-lSOI (N-O.8) Rl-160l f--=-------- -- - -- -- _.__._-_._----_....-_..._---_..._-_ .. _--- - -_._.-.-_._.... _.__..... _------_...... _..__...__ .__..- ----_ *. _'. _..._-------_._._--_._--_. - - - -_._.... __.- ._-_ _- -_._ _-_ _ _- _----_ _-
... .... .. ... .. .....
Control Stations: Blakely, GA (NE- RB-02lS IS) Neals Landing, FL RB-07l8 (SSE-18) Dothan, AL (W-IS) RB-12lS Dothan, AL (W -18) RB-12l8 Webb, AL RB-13ll (WNW-ll) Haleburg, AL (N-l2) RB-1612 r=-- - - - - --.- - -- f-..- ............- ......... .... .. .._-*......._--------- -- _ .__._-_ ._-_._----_._----,. _----_ .._-
-_ _._----_.-_...- ,--_._'.._-- _._-_.. ---- .. -..- .*.-.- ..--- .. ---- -.--- --
.. .......... - - . - .- --
Community Station By sector (NNE-4) RC-OlO4 (NE-4) RC-0204 (ENE-4) RC-0304 (E-S) RC-040S (ESE-S) RC-OSOS (SE-S) RC-060S (SSE-3) RC-0703
TABLE 2-1 (SHEET 4 of 7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types Identification and Locations (sector-miles) (S-5) RC-0805 (SSW-4) RC-0904 (SW-5) RC-I005 (WSW-4) RC-l104 (W-4) RC-1204 (WNW-4) RC-1304 (NW-4) RC-1404 (NNW-4) RC-1504 (N-5) RC-1605 Of Special Interest: Nearest Residence RC-lOOl (SW-l.2) tv I City of Ashford, AL RC-II08
- u. (WSW-8.0)
WATERBORNE Aliquots taken with proportional semi-Surface Water continuous sampler, having a minimum Gamma isotopic analysis of each 4 week sampling frequency not exceeding two composite sample. Tritium analysis for hours, collected weekly for 4 week each quarterly composite.
~___ ._______ ...___...___ .._._._ ..___ comp_O_~!t~~_~l!~Lquan~dY.~<?.!!!p2~tes____.__.______. ..__._._ _________._______.__
Indicator Station: Paper Mill, (~3 miles WRl downstream of plant discharge, RM 40)
-ControTSta-tio- n:--+****--*--** *- -*-***-**** * - -*--*-.-----.----.- ---..--.-------.----.- . . ......-.--..----.-.----.---.---..-..-.-..-..---...- -...--. ------.--..---- --.---.- ----.. - -----.........-..- - _. - --
Upstream of WRB Andrews Lock and dam (~3 miles upstream of the plant intake, RM 47)
TABLE 2-1 (SHEET 5 of7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types Identification and Locations (sector-miles) .. Offsite Ground Water Grab sample quarterly Gamma isotopic, 1-131, and tritium
-- --.--- ..-...*--.... ..-.-----.---- -
~ - -.-.-.-------..- --. ~~~!~~l'~L~_~~~~~Q!2kg~~rterly __________
Indicator Station: Paper Mill Well WG1-07 (SSE-4)
-- ...__ .-._-_._...*._. __ .. _.... _--_......__ __.._---_._._-_._-- _..__.-_ _.......__._-...._ --.- _._..
.-. ..* ... _ *
- _ _ ** _ . . . .. . . _ _ ** _ ** _ _ _ *
- _ _ ___
- _ _ _ ** _ _ . _ ** _ * * *___ ***** w _ _ _ **,._'w _* ___ * ..*.. ___
ControlStation: Whatley Residence WGB-I0 Well (SW-1.2) Onsite Ground Water See Table 2-2 Quarterly sample; pump used to sample GW Tritium, gamma isotopic, and field wells; grab sample from yard drains and parameters (pH, temperature, ponds conductivity, dissolved oxygen, oxidation/reduction potential, and turbidity) of each sample quarterly; Hard to detect radionuclides as necessary based on results of tritium and gamma River Sediment Grab sample semiannually Gamma isotopic analysis of each sample semiannually -.- ._-_.._-_. __ ._-_._---_.__._ .._------- - - -_._--------_.-._-----" ------_. __._- -- - ---_ _ -_._------ --_. __._.- -- .. Indicator Station: Downstream of plant RS1 discharge at Smith's Bend (RM 41t -_._ -_._-- . .. __.-.._.. _ .. - .- __
..........-.__.... -.....*. ._ -._._._ .... __._--- --_.._-_..._..._._._--___..._---_._--r--.-----..--- .--------
.. ------ - ---.- ..- - .---.--.-.-.- -
Control Station: Upstream of plant RSB discharge at Andrews Lock & Dam Reservoir (RM 48)3
TABLE 2-1 (SHEET 6of7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sample Sampling and Collection Frequency Type and Frequency of Analysis with Sample Types Identification and Locations (sector-miles) INGESTION Milk Grab sample bimonthly Gamma isotopic and I-l31 analyses of f---=--........... ....... ---.-... -.... -.---------.-...- .-.. each sa}TIple pimonthly ....-------- - . Control Station: Robert Weir Dairy MB-07l4 NOTE: Samples were no longer available at Donaldsonville, GA this location in 2010_ No replacement (SSE - 14) location has been identified_ Fish Grab sample semiannually for Game Fish Gamma isotopic analysis on the edible and Botto~ Feedi!!S Fish _portions of each sal!!Ple semiannu'!!.!L_ fudic ator-Statlons :---- --- 1-"- '--" - Downstream of plant FGI & FBI discharge in vicinity of Smith's Bend
.. .__._---
.LRM 41)b -, ,_._-_._--_._--_ ,._-----
.... ... ~-.--
_. ... _- --- Control Station: Upstream of plant FGB & FBB discharge in Andrews Lock & Dam Reservoir (RM 4S)b Forage Grab sample monthly_ Gamma isotopic analysis of each sample
-_.._
-_ ....._--..__.__._-_ ._--_._-- _.__._..- monthly_
Indicator Station: South Southeast FI-0701 Perimeter (SSE-l.O) North Perimeter FI-1601 __Q'J"-=.Q_:~l . . _____ _______ .__. I - - --1--- - - _ .._- - . - - -. -.-. Control Station: Dothan, AL (W -IS) FB-121S
TABLE 2-1 (SHEET 7 of 7)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NOTATIONS
- a. These collections are nomlally made at river mile 41.3 for the indicator station and river mile 47.8 for the control station; however, due to river bottom sediment shifting caused by high flows, dredging, etc., collections may be made from river mile 40 to 42 for the indicator station and from river mile 47 to 49 for the control station.
- b. Since a few miles of river water may be needed to obtain adequate fish samples, these river mile positions represent the approximate locations about which the catches are taken. Collections for the indicator station should be from river mile 37.5 to 42.5 and for the control station from river mile 47 to 52.
TABLE 2-2 Onsite Groundwater Monitoring Locations WELL ACQUIFER MONITORING PURPOSE RI Major Shallow Dilution line aquifer R2 Major Shallow Dilution line aquifer R3 Major Shallow Unit 2 RWST aquifer R4 Major Shallow Unit I RWST aquifer R5 Major Shallow Dilution line aquifer R6 Major Shallow Dilution line aquifer R7 Major Shallow Dilution line aquifer R8 Major Shallow Dilution line aquifer R9 Major Shallow Dilution line aquifer RIO Major Shallow Dilution line aquifer Ril Major Shallow Background 1 aquifer R13 Major Shallow Dilution line aquifer R14 Major Shallow Background 2 aquifer PW#2 Drinking water Production Well #2 Supply PW#3 Drinking water Production Well #3 Supply PW#4 Drinking water Production Well #4 Supply CWWest Drinking water Construction Well West Supply CW East Drinking water Construction Well East Supply FRW Drinking water FiringRange Well Supply SW-l N/A Background 3 Service Water Pond East YD N/A Plant outfall East Yard Drain SEYD N/A Plant outfall Southeast Yard Drain 2-9
\
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w E 0801 _1
- -------_.
Radiological Environmental Sampling Locations Indicator Control Community REMP Stations Near the TLD A Plant Perimeter Other *
- Figure 2-1
~ _ .TLD ~ ~~her _ _. __ ~ _ @ @ 2-10
- . -*..*. ------.- - --.- -- -- ~ .
. .
. 134 w 0405 .-A E Radiological Environmental Sampling Locations REMP Stations 2 to 5 TlD .
Indicator
.
Control Community
.t.. Miles From the Plant Other TLD & Other
*
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*
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- Figure 2-2
-_ .. -_.-.- ---.---- ---- .. 2-11
Radiological Environmental Sampling Locations Indicator Control Community REM P Stations Beyond TLD ... ... ... 5 Miles From the Plant Other TLD & Other
*
@ Figure 2-3 2-12
- P'W#2 CW-East R11 N CW-West
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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, community 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 those listed in Table 2-1 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 often detected in REMP samples. It is occasionally detected in the plant's liquid and gaseous effluents. When it is detected in effluents, it is also included in the REMP results. In 2012, Be-7 was detected in liquid effluents at Farley. 3-1
TABLE 3-1 (SHEET 1 of 6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or Type and Minimum Indicator Location with the Highest Community Control Pathway Sampled Total Detectable Locations Annual Mean Locations Locations (Unit of Number of Concentration Mean (b), Mean (b), Mean(b), Measurement) Analyses (MDC) (a) Range N arne Distance Mean (b), Range Range Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 18.0 PC-1101 19.4 18.9 17.3 Particulates 459 2.8-43.4 Plant Entrance 5.6-37.1 4.4-43.3 3.2-50.3 (fCi/m3) (205/205) 0.9 miles (49/49) (153/153) (104/104) WSW
-Oamma----- - ..... .-.--- - - --.---- - - .-.-.-- .-.. -- -..-.- .. - .. .. - --~.-. ---- ....----.---- -.-- l--.---.----...--.--.------------ . -
Isotopic 36 Be-7 24 73.2 PI-0701 83.2 68.5 72.0 47.0-109.5 S. Perimeter 66.1-109.5 53.0-89.1 55.0-100.9 (16/16) 1.0 miles SSE (4/4) (12/12) (8/8) W 1-131 70 NDM NA NDM NDM NDM I N (0/16) (0/16) (0/12) (0/8) Cs-134 50 NDM(c) NA(d) NDM NDM (0/16) (0/12) (0/8) Cs-137 60 NDM NA NDM NDM (0/16) (0/12) (0/8) Airborne 1-131 70 NDM NA NDM NDM NDM Radioiodine 359 (0/203) (0/52) (0/104) (fCi/m3) Direct Radiation Gamma NA 17.4 RI-0401 25.0 14.7 15.8 (mRl91 days) Dose 12.6-26.5 Pint Perimeter 23.6-26.5 10.8-18.6 1l.1-20.5 159 (64/64) 0.8 miles E (4/4) (71171) (24124) Milk (PCill) Gamma Isotopic 0 Cs-134 15 NA NA NA NA Cs-137 18 NA NA NA NA Ba-140 60 NA NA NA NA La-140
- _15 _ ., __**** ___*__ _ _ _ H._ . ._
NA NA NA NA
.- .----.... - - ..
.. - -- .- .~
.......... -_..-.. _-- .......... -._ .__. * ' . " . . .. R .. .. ___ . _ *** _ . . . _ * * **
- H * * * *
- _ _ _ _ _ _ _
- _ _ _ _ _ ** *** .._____ ,......__.. __......._______ ..._. __ .. _..H._._. ~----- --- ----- _.._._----- -
1-131 1 NA NA NA NA
TABLE 3-1 (SHEET 2 of 6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or Type and ' Minimum I Indicator I Location with the Highest Community Control Pathway Sampled Total Detectable Locations Annual Mean Locations Locations (Unit of Number of Concentrati on Mean (b), Mean (b), Mean(b), Measurement) Analyses (MDC) (a) Range . Name Distance Mean (b), Range Range Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Forage Gamma (PCi/kg wet) Isotopic 39 Be-7 729 1791 FI-160 1 1944 NA 1454 (761-2632) N. Perimeter (761-2505) (454-2486) (24/24) 0.8 miles N (12112) (12112) 1-131 60 NDM NA NA NDM (0/24) (0112) Cs-134 60 NDM NA NA NDM (0124) (0112) w, Cs-137 80 NDM FB-1218 9.44 NA 9.44 w (0/24) Dothan, AL (1112) Offsite Ground H-3 2000 NDM NA NA NDM Water (pCi/l) 8 ...--- ------.-... ...- .
~- -. -. - .. ~-
-_.._._.__._... _-_... _ -- _._. iQL41_________ N'X----- *- *----- __ _ _ * * *
- _ ** _ _ * *
- _ _ w
- _ _
- _ _ *w** __ _ ._** _ _ _ _ _ _ *****. (0/4)
(g) 1-131 1 NDM NA NDM 8----_.._._-- -"- ---.
.*. _ ' *...._...* ..-.--_ _ _- -_._--
, .. ... i2L 41 ___.__ __ _-_.- .._.*. _._-_.._-_ __
....*. __ .. .*.. . -_._--- --_.-_..._ _._-------- (0/4)
Gamma Isotopic 8 Mn-54 15 NDM NA NA NDM (0/4) (0/4) Fe-59 30 NDM NA NA NDM (0/4) (0/4) Co-58 15 NDM NA NA NDM (0/4) (0/4) Co-60 15 NDM NA NA NDM (0/4) (0/4) Zn-65 30 NDM NA NA NDM (0/4) (0/4) Zr-95 30 NDM NA NA NDM (0/4) (0/4)
TABLE 3-1 (SHEET 3 of6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or Type and Minimum Indicator Location with the Highest Community Control Pathway Total Detectable Locations Annual Mean Locations Locations Sampled Number of Concentration Mean (b), Mean (b), Mean(b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Range Range Measurement) Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Nb-95 15 NDM NA NA NDM (0/4) (0/4) Cs-134 15 NDM NA NA NDM (0/4) (0/4) Cs-137 18 NDM NA NA NDM (0/4) (0/4) Ba-140 60 NDM NA NA NDM (0/4) (0/4) La-140 15 NDM NA NA NDM (0/4) (0/4) Surface Water H-3 3000 390 Ga Pacific 390 NA 223 (pCi/l) 8 (3/4) Paper Mill (3/4) (114)
*... _ ---- - _._.*._- -_ .*.._ _.. -
.. ..
RM40
- - --.--.-.-.--- .. -.- ... ------.-. ..*.*
Gamma Isotopic 26 Be-7 124 (e) NDM NA NA NDM (0/13) (0/ 13) Mn-54 15 NDM NA NA NDM (0/13) (0/ 13) Fe-59 30 NDM NA NA NDM (0/ 13) (0/ 13) Co-58 15 NDM NA NA NDM (0/13) (0/ 13) Co-60 15 NDM NA NA NDM (0/ 13) (0/13) Zn-65 30 NDM NA NA NDM (0113) (0/ 13) Zr-95 30 NDM NA NA NDM (0/13) (0/13) Nb-95 15 NDM NA NA NDM (0/ 13) (0/ 13)
TABLE 3-1 (SHEET 4 of6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or Type and Minimum Indicator Location with the Highest Community Control Pathway Total Detectable Locations Annual Mean Locations Locations Sampled Numberof . Concentration Mean (b), Mean (b), Mean(b), (Unit of Analyses (MDC) (a) Range N arne Distance Mean (b), Range Range Measurement) Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) 1-131 15 (f) NDM NA NA NDM (0/13) (0/13) Cs-134 15 NDM NA NA NDM (0/13) (0113) Cs-137 18 NDM NA NA NDM (0/13) (0/13) Ba-140 60 NDM NA NA NDM (0/13) (0/13) La-140 15 NDM NA NA NDM (0/13) (0113) Bottom Gamma Feeding Fish Isotopic (pCi/kg wet) 4 Be-7 655 (e) NDM NA NA NDM (0/2) (012) Mn-54 130 NDM NA NA NDM (012) (0/2) Fe-59 260 NDM NA NA NDM (012) (0/2) Co-58 130 NDM NA NA NDM (0/2) (012) Co-60 130 NDM NA NA NDM (012) (0/2) Zn-65 260 NDM NA NA NDM (012) (0/2) Cs-134 130 NDM NA NA NDM (012) (0/2) Cs-137 150 N/A NA NA NDM (0/2) (012)
TABLE 3-1 (SHEET 5 of 6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or Type and Minimum Indicator Location with the Highest Community Control Pathway Total Detectable Locations Annual Mean Locations Locations Sampled Number of Concentration Mean (b), Mean (b), Mean(b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Range Range Measurement) Performed (Fraction) & Direction Range (Fraction) (Fractiof!) (Fraction) Game Fish Garruna (pCilkg wet) Isotopic 4 Be-7 655 (e) NDM NA NA NDM (0/2) (0/2) Mn-54 130 NDM NA NA NDM (0/2) (0/2) Fe-59 260 NDM NA NA NDM (0/2) (0/2) Co-58 130 NDM NA NA NDM (0/2) (0/2) Co-60 130 NDM NA NA NDM (0/2) (0/2) Zn-65 260 NDM NA NA NDM (0/2) (0/2) Cs-134 130 NDM NA NA NDM (0/2) (0/2) Cs-137 150 15.449 Downstream, 15.449 NA NDM 14.0-16.9 near Smith's 14.0-16.9 (0/2) (2/2) Bend (RM 41) (2/2) River Garruna Shoreline Isotopic Sediment 4 (pCi/kg dry) Be-7 655 (e) NDM NA NA NDM (0/2) (0/2) Cs-134 150 NDM NA NA NDM (0/2) (0/2) Cs-137 180 NDM NA NA NDM (0/2) (0/2)
TABLE 3-1 (SHEET 6 of 6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama NOTATIONS
- 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 (Table 4-1 of this report). 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 parentheses.
- 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 Table 4-1 of this W
report. I -.J
- f. If a drinking water pathway exists, a value of I pCi/1 would be used. See note b of Table 4-1 of this report.
- g. On site groundwater results are discussed in Section 4.6.
4.0 DISCUSSION OF RESULTS Included in this section are evaluations of the laboratory results for the various sample types. The Minimum Detectable Difference compares the lowest significant difference between a control station and an indicator station, or the control station and the community station, that can be detennined statistically at the 99% Confidence Level (CL). The MDD was detennined using the standard Student's t-test. MDD as a tool can quantify plant Farley's impact on the surrounding environment. A difference in the mean values which was less than the MDD was considered to be statistically indiscernible. The 2012 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 O's and listed in the tables as NDM. Table 4-1 Minimum Detectable Concentrations (MDC) Analysis Water Airborne Fish Milk Grass or Sediment (pCill) Particulate (pCilkg) (pCill) Leafy (pCilkg) or Gases wet Vegetation dry (fCilm3) (pCilkg) 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 (a) Ifno drinking water pathway exists, a value of 3000 pCi/l may be used. 4-1
(b) If no drinking water pathway exists, a value of 15 pCi/1 may be used. Table 4-2 Reporting Levels (RL) Analysis Water Airborne Fish Milk (pCUl) Grass or (pCUJ) Particulate (pCUkg) wet Leafy or Gases Vegetation (fCUm3) (pCUkg) 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 8a-140 200 300 La-140 100 400 (a) This is the 40 CFR 141 value for drinking water samples. If no drinking water pathway exists, a value of 30,000 may be used. (b) If no drinking water pathway exists, a value of 20 pCi/1 may be used. Atmospheric nuclear weapons tests from the mid 1940's through 1980 distributed man-made nuclides around the world. The most recent atmospheric tests in the 1970's 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. Significant upward trends also followed the Chernobyl incident, which began on April 26, 1986. The most significant nuclear event since Chernobyl occurred at Fukushima Daiichi Nuclear Power Plant after the Tohoku earthquake and tsunami on March 11 , 2011. Equipment failures and nuclear meltdowns resulted in radioactivity being released into the atmosphere. Southern Nuclear's three sites (Farley, Hatch, and Vogtle) detected 1-131 in REMP samples for several weeks following the disaster. 4-2
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. 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 vialues which differed from the mean of a set by a statistically significant amount. Identified outliers were investigated to determine the reason(s) for the variation. 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 1) DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLE(S) 12/ 1311 0 - 03/21112 PI-07011II-0701 No flow measurement of totalizer Flow turbine not operating Sample flow verified adequate; 48 CR 649 I 9rrE 400241 1.0 mile SSE display correctly; no flow indicated on LPM used as estimating sample totalizer flow for the affected collection weeks of2010, 2011, and 2012. 03/ 14/ 12 to 03/16/ 12 PI-1601111-1601 Annual preventive maintenance and Work management Work Order SNC93410 CR 423799rrE 331708 0.8 miles N calibration of station's sampling coordinatiOn/scheduling rescheduled; station operation equipment went beyond the late date found satisfactory upon performance of required tasks. 03 /20/12 - 04/10/12 PC-1108IIC- Non-representative sample of Partial loss of sample from leak on Pump inlet seal tightened to stop CR 430318rrE 355081 1108 airborne particulates inlet seal on vacuum pump. Work air leakage around pump. 8.0 miles WSW Order SN C3 81864 submitted. 03-20 03/27/12 PI-160 1111-1601 Non-representative sample of Partial loss of sample from leak on Pump inlet seal tightened to stop CR 430325TE 355082 0.8 mile N airborne particulates inlet seal on vacuum pump. Work air leakage around pump order SNC381866 submitted. I s, quarter 2012 OSLD Station OSLD missing from station OSLD packet lost from retaining Replaced TLDs at beginning of CR 434999TE 360851 RC-1504A&B clip, search of area did not locate quarter 4.0 miles NNW missing OSLD. 06/11112 - 6/19/12 PI-0701 111-070 1 Non-representative sample of Sample station lost power for Station operation satisfactory after CR 472380rrE 438787 1.0 mile SSE airborne particulates approximately 4.4 hours due to power restored inoperable electrical transformer supplying power to sampler. 07/24/12 through end of PI-I 101111-1 101 No flow measurement of totalizer Flow turbine not operating Sample flow verified adequate; 48 year 0.9 miles WSW display correctly; no flow indicated on LPM used as estimating sample CR 489640/TE 480724 totalizer flow for remaining collection weeks of2012. 08/14/12 - 09/20/12 PI-160 1111-160 1 Non-representative sample of Lost approximately 37.1 days of Station operation satisfactory after CR 523761rrE 524246 0.8 miles N airborne particulates sample time; cause was electrical power restored power supply issue due to lightning strike 10/23/12 to 10/30/12 PB-1218/IB- Air volume totalizer found out-of- Found mid-range input at 3.4 cfm, Recalibration successful CR 539917rrE 540412 1218 tolerance high upon calibration (tolerance range 1.8 - 2.2 cfm.) 18 miles W
4.1 Land Use Census In accordance with ODCM 4.1.2, a land use census was conducted during the month of November 2012. The land use census is used to determine the locations of the nearest permanent residence and milk animal in each of the 16 compass sectors within a distance of 5 miles. A milk animal is a cow or goat producing milk for human consumption. The 2012 survey revealed no significant changes from the 2011 survey. No milk animals were found within a 5 mile distance. 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 ANIMAL N 2.6 none NNE 2.5 none NE 2.4 none ENE 2.4 none E 2.8 none ESE 3.0 none SE 3.4 none SSE >5 none S 4.3 none SSW 2.9 none SW 1.2 none WSW 2.4 none W 1.3 none WNW 2.1 none NW 1.5 none NNW 3.4 none The Houston County, Alabama and the Early County, Georgia Extension Agents were contacted for assistance in locating commercial dairy farms and privately owned milk animals within 5 mi les of the plant. A list of commercial dairy farms in Houston County, AL and Seminole County, GA was provided; there are no commercial dairy farms in Early County, GA. Neither agent knew of privately owned milk animals within 5 miles of FNP. In addition, field surveys were conducted in the plant vicinity along the state and county highways and the interconnecting secondary roads. No milk animals were found within 5 miles of the plant. ODCM 4.1.2.2.1 requires a new controlling receptor to be determined, if the land use census identifies a location that yields a calculated receptor dose greater than the one in current use. Neither current sampling locations nor the controlling receptor were affected by the 2012 land use census results. The current controlling receptor as described in ODCM Table 3-7 remains a child in the SW Sector at 1.2 miles. 4-5
4.2 Airborne As specified in Table 2-1 and shown in Figures 4.2-1 and 4.2-2, airborne particulate filters and charcoal canisters are collected weekly at 4 indicator, 3 control, and 3 community stations. Particulate filters are collected at all of the stations while the charcoal canisters are collected at all but 2 of the community stations. At each location, air is continuously drawn through a glass fiber filter to retain airborne particulates and, as appropriate, an activated charcoal canister is placed in series 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 2012 annual average weekly gross beta activity was 18.0 fCi/m3 at the indicator stations and 17.3 fCi/m 3 at the control stations. The difference of 0.7 fCi/m 3 between the two averages is statistically indiscernible since it is less than the MDD of 2.6 fCi/m 3. The 2012 annual average weekly gross beta concentration was 18.9 fCi/m 3 at the community stations. The community stations average was 1.6 fCi/m 3 greater than the average for the control stations. The difference is statistically indiscernible since it is less than the MDD of 2.7 fCi/m3. During calibration of one of the control stations (PB-1218) in October, 2012, the mid-range input for the air volume totalizer was found out of tolerance high. The totalizer was re-calibrated successfully. Gross beta sample results from this sample station showed an increase for the last two months of the year compared to the January through October results. However, this increase was consistent with the results of the other air sampling stations when their results were compared for the same time periods. Due to the weapons tests during preoperation and the early years of operation, the average gross beta concentrations were 5 to 10 times greater than those currently being measured. By the mid 1980s, the readings had diminished to about half the current levels. These annual averages approximately doubled as a consequence of the Chernobyl incident in 1986; this impact faded away in approximately 2 years. The installation of new air monitoring equipment in 1992 yielded an approximate factor of 2 increase in the readings. Since then, the levels have not varied significantly. The nuclear accident at Fukushima Daiichi Nuclear Power Plant which occurred after the Tohoku earthquake and tsunami on March 11, 2011 released radioactivity into the environment that was detected in Farley air samples. Iodine-131 was detected in air cartridges after Fukushima but no changes in gross beta activity were seen during that same time period. The historical trending of the average weekly gross beta air concentrations for each year of operation and the preoperational period at the indicator, control and community stations is plotted in Figure 4 .2-1 and listed in Table 4.2-1. Historical values for the mean gross beta and values for 2012 support the position that the plant's contribution to gross beta concentration in air is insignificant. 4-6
Figure 4.2-1 Average Weekly Gross Beta Air Concentration 250 200 M E 150 1\
g \ ... J c: o 100 ,- -
~
c: Q) g o (J 50 \ I1\ I-' /r-,
.,
"""'1 k .<tfJ I I j I o
~ n ~ ~ M ~ ~ 00 ~ ~ % ~ 0 2 4 6 8 10 12 Year
- -MOC -Indicator ----- Control - - Community 4-7
Table 4.2-1 Average Weekly Gross Beta Air Concentration Period Indicator Control Community (fCilm3) .(fCilm3) (fCilm3) Pre-oQ 90 92 91 1977 205 206 206 1978 125 115 115 1979 27.3 27.3 28.7 1980 29.7 28.1 29.2 1981 121 115 115 1982 20.0 20.4 21.0 1983 15.5 14.1 14.5 1984 10.2 12.6 10.5 1985 9.0 9.6 10.3 1986 10.5 15.8 12.5 1987 9.0 11.0 17.0 1988 8.0 8.0 10.0 1989 7.0 7.0 8.0 1990 10.0 10.0 10.0 1991 9.0 10.0 8.0 1992 15.0 17.9 18.5 1993 19.1 22.3 22.4 1994 19.0 20.0 19.0 1995 21.7 22.9 21.6 1996 20.3 22.3 23.5 1997 21.1 21.6 22.4 1998 20.6 19.3 22.0 1999 20.5 22.1 25.2 2000 20.9 20.8 23 .6 2001 16.3 17.2 17.3 2002 16.8 18.0 16.8 2003 19.1 19.3 19.9 2004 22.0 21.3 22.4 2005 18.4 19.3 19.0 2006 16.1 17.5 16.8 2007 14.5 18.9 17.3 2008 16.7 20.6 18.0 2009 16.2 16.3 17.3 2010 21.2 17.5 18.2 2011 20.9 14.5 18.2 2012 18.0 17.3 18.9 4-8
During 2012, Be-7 was the only radioisotope detected from the gamma isotopic analysis of the quarterly composites of the air particulate filters. This has generally been the case since the impact of the weapons tests and the Chemobyl incident have faded. Be-7 is naturally occurring but also was detected in Farley's gaseous effluents. The average Be-7 at the indicator stations was 73.2 fCi/m 3 and the average at the control stations was 72.0 fCi/m 3 . The difference of 1.2 fCi/m 3 is statistically indiscernible when compared with the MDD of 21.0 fCilm 3 . The average Be-7 at the community stations was 68.5 fCi/m 3 . The difference (3.5 fCi/m 3) between the control and community stations is not statistically discernible since it is less than the MDD of 18.1 fCi/m 3 . There is no required MDC or Reporting Level for Be-7. During preoperation and the early years of operation, a number of fission and activation products were detected. During preoperation, the average levels for Cs-134 and Cs-137 were 22 and 9 fCi/m 3, respectively. In 1986, as a consequence of the Chemobyl incident, Cs-134 and Cs-137 levels of 3 to 4 fCi/m 3 were found. The MDC and RL for Cs-134 are 50 and 10,000 fCi/m 3 and the MDC and RL for Cs-137 are 60 and 20,000 fCi/m 3 respectively. The historical trending of the annual detectable Cs-137 concentrations for the indicator, control and community stations is provided in Figure 4.2-2 and Table 4.2-2. The trend has been generally downward since preoperation and no positive results have been observed since 1988. 4-9
Figure 4.2-2 Average Annual Cs-137 Concentration in Air 20 M 15 ~ g I: o 10 -
- .;:;
~
I: CD CJ I: o 5
\
j
~~~~
U
~ be(
~
~ \ ./~,
o 8 10 12 Year
--Indicator - Control - - Community I 4-10
Table 4.2-2 Average Annual Cs-137 Concentration in Air Period Indicator Control Community (fCilm3) (fCilm3) (fCilm3) Pre-op 8 13 7 1977 3.0 3.0 3.0 1978 4.0 5.0 5.0 1979 2.0 NOM 2.0 1980 1.0 2.0 1.8 1981 2.8 3.2 2.6 1982 1.7 NOM 1.0 1983 1.0 NOM 1.0 1984 NOM 1.5 NDM 1985 1.0 1.0 1.0 1986 3.3 3.4 2.7 1987 NDM NDM NOM 1988 NOM NOM 1 1989 NDM NOM NOM 1990 NDM NOM NOM 1991 NOM NOM NOM 1992 NDM NOM NOM 1993 NOM NDM NDM 1994 NOM NOM NOM 1995 NOM NDM NOM 1996 NOM NDM NOM 1997 NOM NOM NOM 1998 NDM NOM NOM 1999 NOM NDM NOM 2000 NDM NDM NOM 2001 NOM NOM NDM 2002 NOM NOM NOM 2003 NOM NOM NOM 2004 NOM NOM NDM 2005 NDM NOM NDM 2006 NOM NOM NDM 2007 NDM NOM NOM 2008 NOM NDM NOM 2009 NOM NOM NOM 2010 NOM NDM NDM 2011 NDM NOM NOM 2012 NOM NOM NOM 4-11
Airborne I-131 was not detected in charcoal canister samples in 2012. As discussed earlier in this section, I-131 activity was detected in 2011 (ranging from 32.5 to 115.0 fCi/m 3 ) and was attributed to the Fukushima nuclear accident. In 1978, levels between 40 and 50 fCi/m 3 were found in a few samples and attributed to the Chinese weapons tests; then after the Chernobyl incident, levels up to a few hundred fCi/m 3 were found in some samples. At no other times has airborne I-131 been detected in the environmental samples. The MDC and RL for airborne 1-131 are 70 and 900 fCilm) respectively. Table 4-3 lists REMP deviations that occurred during 2012. There were eight sampling deviations related to air sampling listed in Table 4-3. Two issues with the sampler flow totalizer display resulted in estimated flow rates being used after adequate flow was verified. Two deviations were related to loss of power at the sampling stations, one of short duration not affecting collection of a valid sample, and one of extended duration that resulted in no sample for several weeks while repairs were made. Two deviations were related to air sampler calibrations and did not affect sample collection. Two issues were identified with partial loss of sample flow due to vacuum pump seal leaks; these issues resulted in non-representative samples. 4-12
4.3 Direct Radiation In 2012, direct (external) radiation was measured with Landauer InLight optically stimulated luminescent (OSL) dosimeters, which replaced the Panasonic thermo luminescent dosimeters (TLDs). Two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges. Direct radiation stations are established in each of the 16 sectors, to form 2 concentric rings. The inner ring stations are located near the plant perimeter, as shown in Figure 2-1, and the outer ring stations are located at distances of approximately 3 to 5 miles from the plant, as shown in Figure 2-2. The stations forming the inner ring are designated as the indicator stations. The 6 control stations are located at distances greater than 10 miles from the plant, as shown in Figure 2-3 . Stations are also provided which monitor special interest areas: the nearest occupied residence (SWat 1.2 miles), as shown in Figure 2-1, and the city of Ashford (WSW at 8 miles), as shown in Figure 2-3. The 16 outer ring stations and the 2 special interest stations are designated as community stations. As provided in Table 3-1, the average quarterly exposure measured at the indicator stations (inner ring) during 2012 was 17.4 mR which was 1.6 mR greater than the 15.8 mR measured at the control stations. This difference is not statistically discernible since it is less than the MDD of 1.9 mR. The difference of 1.1 mR found between the control stations (15.8 mR) and community stations (14.7 mR) is not statistically discernible since the difference is less than the MDD of 1.2 mR. The difference between the indicator and control and between the control and community stations is consistent with what has been seen in previous years. The historical trending of the average quarterly exposures in units of mR at the indicator, control, and community locations are plotted in Figure 4.3-1 and listed in Table 4.3-1. During preoperation the average exposure at the indicator stations was 1.2 mR greater than that for the control stations, compared to the average over the entire period of operation which was 1.3 mR greater. During preoperation, the average exposure at the control stations was 1.3 mR greater than that at the community stations and the average over the period of operation was 1.3 mR greater. This supports the position that the plant is not contributing significantly to direct radiation in the environment. 4-13
During 20 I 0, OSL badges were co-located on station with the TLD badges. In 2011, only the OSL badges were placed at each station. Following the change to only OSL badges, the differences between indicator, control, and community locations has been consistent with previous years. An increase noted in 2010 reflects issues (especially during 2 nd Qtr) with the aging Panasonic TLD reader. Table 4-3 lists the REMP program deviations that occurred in 2012 . There was one sample deviation involving OSL badges. Both RC-1S04A & B were missing at the sample station at the time of the first quarter badge collection and second quarter OSL placement. The new badges were placed with the other second quarter OSLs. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 30 25 j / ~ t .,~1\~ ~ ~ 20 -1
,
11\ v §. \ I.-"' ~
'"--,
... \- /. ~ ~ ~
~V
'~
Q)
'-., V
~
- I 15 t- -. V' 1/1 It V 'I \, / \~ ---......, ~ V
- ........, v ....... ~
Vl 0 Co )( W 10 ~ " '"----1 ~ V ~ V" N R~ ~ 5 0 , , , " , Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year C --Indlcator - Control - Community 4-14
Table 4.3-1 Average Quarterly Exposure from Direct Radiation Period Indicator (mR) Control (mR) Community (mR) Pre-op 12.6 11.4 10.1 1977 10.6 12.2 10.6 1978 15.0 13.5 12.0 1979 20.3 18.7 15.2 1980 21.9 21.6 18.5 1981 16.5 14.9 14.5 1982 15.5 14.7 13.0 1983 20.2 20.2 17.4 1984 18.3 16.9 15.3 1985 21.9 22.0 18.0 1986 17.8 17.7 15.1 1987 20.8 20.0 18.0 1988 21.5 19.9 18.5 1989 18.0 16.2 15.3 1990 18.9 16.4 15.8 1991 18.4 16.1 16.1 1992 16.1 13.6 13.5 1993 17.4 15.9 15.6 1994 15 .0 13.0 12.0 1995 14.0 12.5 11.8 1996 14.2 12.7 11.9 1997 15.3 13.9 11.9 1998 16.2 14.6 13.9 1999 14.7 13.4 12.6 2000 15.5 14.1 13.5 2001 14.9 13.4 12.7 2002 14.1 12.6 11.9 2003 15.2 13.6 12.9 2004 14.3 12.9 12.1 2005 14.7 13.4 12.5 2006 15.2 13.6 12.9 2007 14.6 13.3 12.5 2008 15.0 13.7 12.9 2009 15.2 13.6 12.8 2010 17.8 16.7 15.5 2011 15.1 14.4 13.0 2012 17.4 15.8 14.7 4-15
The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self imposed limit of 3.5. Previously with TLDs, this limit had been 1.4. However, the OSL readings varied more (between the two elements) than the TLD readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication 15D, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 2010. This limit serves as a flag to initiate an investigation, To be conservative, readings with a standard deviation greater than or equal to 3.5 are excluded since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 2012, the following OSL results were excluded from the data set because their standard deviations were greater than or equal to 3.5: First Quarter None Second Quarter None Third Quarter None Fourth Quarter None No badges at any station exhibited a standard deviation greater than or equal to 3.5 in 2012. 4-16
4.4 Milk Milk samples had been collected biweekly from a control location until the end of 2009 when the dairy would no longer provide samples. No indicator station (a location within five miles of the plant) has been available for milk sampling since 1987. As discussed in Section 4.0, no milk animals were found within five miles of the plant during the 2012 land use census therefore no milk sampling was perfonned during 2012. Per Table 2.1, gamma isotopic analyses were perfonned on milk samples when they were collected in previous years. Cs-137 and 1-131 are the only man-made radionuclides that have been identified over the history of milk sampling. The historical trending of the average annual detectable Cs-137 concentration in milk samples is shown in Figure 4.4-1 and Table 4.4-1. Cs-137 has not been detected in milk since 1986. Its presence at that time is attributed to the Chemobyl incident. The earlier detectable results were attributed to weapons testing. The MDC and RL for Cs-137 in milk are 18 and 70 pCi/I, respectively. Figure 4.4-1 Average Annual Cs-137 Concentration in Milk 45 i
;
40 i
/ -
U 35 30
'rI -Co I:
0 25
; -...
10 I: Q) CJ 20 15
/"\.
\ I ~,
\
I: U 0 10 \ \ /\ 5 \\ \ I\ i I II \ 0 Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 Year I -+-Indlcator _Control --MOC 4-17
Table 4.4-1 Average Annual Cs-137 Concentration in Milk Period Indicator Control (pCi/l) (pCi/l) Pre-op 32 18 1977 41 20 1978 15 17 1979 NOM NOM 1980 NOM NOM 1981 NOM 23.0 1982 NOM NOM 1983 NOM NOM 1984 NOM NOM 1985 NDM NOM 1986 NOM 16.5 1987 NOM NOM 1988 NOM NOM 1989 NDM NOM 1990 NOM NDM 1991 NOM NDM 1992 NOM NOM 1993 NOM NOM 1994 NOM NOM 1995 NOM NOM 1996 NOM NOM 1997 NOM NOM 1998 NOM NOM 1999 NOM NOM 2000 NOM NOM 2001 NOM NOM 2002 NOM NDM 2003 NDM NDM 2004 NOM NDM 2005 NOM NOM 2006 NDM NOM 2007 NOM NDM 2008 NDM NOM 2009 NOM NDM 2010 No sampJe No sample 2011 No sample No sample 2012 No sample No salll~e 4-18
As specified in Table 2-1, milk samples were also analyzed for 1-131, which has not been detected in milk since 1986. The presence of 1-131 at that time is attributed to the Chernobyl incident. The earlier detectable results were attributed to weapons testing. The MDC and RL for 1-131 are 1 and 3 pCi/l, respectively. Figure 4.4-2 and Table 4.4-2 show the historical trending of the average annual detectable 1-131 concentration in milk samples. Figure 4.4-2 Average Annual 1-131 Concentration in Milk 50 40 - S U - a. c: 30 1\ - \/ 0
...
CI:I c: (I) 20 u c: 0 U 10 1\
\ /\ I /\ I - - I o
Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 Year
---Indicator - - Control - MOC RL I 4-19
Table 4.4-2 Average Annual 1-131 Concentration in Milk Period Indicator Control (pCiJl) (pCiJI) Pre-op 41 14 1977 20 2.6 1978 30 11 1979 NDM NDM 1980 NOM NDM 1981 NDM NOM 1982 NOM NDM 1983 NDM NDM 1984 NDM NDM 1985 NDM NOM 1986 NOM 5.0 1987 NDM NOM 1988 NDM NDM 1989 NOM NOM 1990 NDM NDM 1991 NDM NOM 1992 NDM NDM 1993 NOM NOM 1994 NDM NDM 1995 NOM NDM 1996 NDM NOM 1997 NOM NDM 1998 NDM NDM 1999 NDM NDM 2000 NDM NOM 2001 NDM NDM 2002 NDM NDM 2003 NDM NDM 2004 NDM NDM 2005 NDM NDM 2006 NDM NDM 2007 NDM NDM 2008 NOM NDM 2009 NOM NDM 2010 No sample No sample 2011 No sample No sample 2012 No sample No sam~le 4-20
4.5 Forage In accordance with Table 2-1, forage samples are collected every 4 weeks at two indicator stations on the plant perimeter, and at one control station located approximately 18 miles west of the plant, in Dothan. Gamma isotopic analyses are perfonned on each sample. During preoperation and the years of operation through 1986 (the year of the Chernobyl incident), Cs-137 was typically found in about a third of the 35 to 40 forage samples collected per year. In 1987 and 1988 the number dropped to about a seventh of the total samples and from 1989 through 1994, it was only found in one or two samples every year. From 1994 to 2006, Cs-13 7 was detected in only a few samples, three indicator samples and three control samples. In 2012, Cs-137 was detected in one of the 12 control samples, at 9.44 pCi/kg wet, and in none of the 24 indicator samples. The occasional presence of Cs-13 7 in vegetation samples is attributed primarily to fallout from nuclear weapons tests and from the Chernoby1 incident. The MDC and RL for Cs-137 in forage are 80 and 2000 pCi/kg wet, respectively. Table 4.5-1 presents the average detectable results of Cs-137 found in forage over the life of the plant and Figure 4.5-1 shows the historical trending of this data. Be-7 is naturally occurring but was also detected in liquid and gaseous effluent samples in 2012 therefore it was reported when detected in REMP samples in 2012. All forage indicator and control samples were positive for Be-7. The average Be-7 at the indicator stations was 1790 pCi/kg wet and the average at the control station was 1453 pCi/kg wet. The difference of 337 pei/kg wet is less than the MDD of 500 pCi/kg and therefore is not statistically discernible. There is no Required MDC or Reporting Level for Be-7. 4-21
Figure 4.5-1 Average Annual Cs-137 Concentration in Forage 90 80 I j\ V} II II I'-' ~ K
\ ~, 'v 1\
~YI \ ,1\
\ \ \
I 1\ N II ~
\"--,
V 10 I 1\ I \ I \
\!I h ,-- .\ / \ IV1\
o II \ 1/ ' I \ I \1/\ \ V.\ !/ Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year
--Indicator - Control - MDC 4-22
Table 4.5-1 Average Annual Cs-137 Concentration in Forage Period Indicator Control (pCilkgl wet (pCilkg) wet Pre-op 59.4 48 .6 1977 25.0 NDM 1978 52.5 32.5 1979 37.2 32.8 1980 36.2 35.9 1981 32.1 43.1 1982 25.0 33.1 1983 16.8 23.6 1984 19.9 22.8 1985 22.2 9.5 1986 41.2 36.2 1987 46.8 NDM 1988 33.6 31.7 1989 35.7 NDM 1990 56.0 NDM 1991 NDM 12.9 1992 NDM 43.0 1993 NDM 24.0 1994 NDM 24 1995 NDM NDM 1996 NDM NDM 1997 52.6 NDM 1998 NDM 22.7 1999 NDM NDM 2000 NDM NDM 2001 NDM NDM 2002 NDM NDM 2003 24.1 25 .2 2004 2l.6 NDM 2005 NDM 23.1 2006 NDM NDM 2007 NDM NDM 2008 10.1 NDM 2009 NDM NDM 2010 NDM NDM 2011 NDM NDM 2012 NDM 9.4 4-23
During preoperation and in the early years of operation, 1-131 was found in 10% to 25% of the forage samples at very high levels which ranged from around 100 to 1,000 pCi/kg wet. In 1986 (Chernobyl incident), 1-131 reappeared after not having been detected for 3 years. In March 2011, after the nuclear accident at Fukushima Daiichi Nuclear Power Plant, Southern Nuclear's three sites (Farley, Hatch, and Vogtle) detected 1-131 in REMP samples for several weeks following the disaster. Iodine-131 was detected in one forage sample (20.4 pCi/kg wet) at Farley (the North Perimeter indicator station) after the Fukushima event. In 2012, 1-131 was not detected in any of the forage samples at the indicator or control locations. The MDC and RL for 1-131 are 60 and 100 pCi/kg wet, respectively. Table 4.5-2 lists the average detectable results of 1-131 found in forage over the life of the plant and Figure 4.5-2 plots the historical trending of this data. Figure 4.5-2 Average Annual 1-131 Concentration in Forage 1000 900
!
_ 800 I Q)
~
Ci 700 - ~ U 600 -g c. 500 1/1\
ctI ~ 400 c: Q) g 300 o U 200 I 100 o
!\, ~~ 1
/I\i\L i ! I Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year I ---Indicator ___ Control -MOC - RL I 4-24
Table 4.5-2 Average Annual 1-131 Concentration in Forage Period Indicator Control (pCilk2) wet (pCilkg) wet Pre-op 405 486 1977 971 654 1978 220 240 1979 NDM NDM 1980 NOM NDM 1981 21.4 NOM 1982 46.4 NDM 1983 NDM NOM 1984 NDM NDM 1985 NOM NOM 1986 184 NDM 1987 NOM NDM 1988 NOM NDM 1989 NDM NDM 1990 NDM NOM 1991 NDM NDM 1992 NDM NDM 1993 NDM NDM 1994 NOM NDM 1995 NDM NOM 1996 NDM NDM 1997 NDM NDM 1998 NDM NDM 1999 NDM NOM 2000 NDM NDM 2001 NDM NOM 2002 NDM NOM 2003 NDM NDM 2004 NDM NDM 2005 NDM NOM 2006 NDM NDM 2007 NDM NDM 2008 NDM NDM 2009 NDM NDM 2010 NDM NDM 2011 20.4 NDM 2012 NDM NDM 4-25
These forage analyses results show the impact of the weapons tests during preoperation and the early years of operation and of the Chemobyl incident in 1986 and for a few years afterwards. The impact is reflected by the number of different radionuclides detected, the fraction of samples with detectable results, as well as the magnitude of the results. During preoperation and for the first few years of operation, 11 different radionuclides from fission and activation products were detected. By 1985, only 2 different radionuclides were detected and the fraction of samples with detectable results had diminished. In 1986, the same two nuclides as seen in 1985 appeared at a significantly higher magnitude and 1-131 reappeared. In the years following 1986, only Cs-137 has been found in forage and it has been found in a decreasing fraction of the samples. 4-26
4.6 Ground Water In the FNP offsite environs, there are no true indicator sources of ground water. A well, located about four miles south-southeast of the plant on the east bank of the Chattahoochee River, serves Georgia Pacific Paper Company as a source of potable water and is designated as the indicator station. A deep well located about 1.2 miles southwest of the plant, which supplies water to the Whatley residence, is designated as the control station. Samples are collected quarterly and analyzed for gamma isotopic, 1-131 and tritium as specified in Table 2-1. In 2012, there were no radionuclides detected in any of the ground water samples from either sample station. In 1983, 1985, and 1986, Cs-134 was detected in single samples at levels ranging from 3 to 13 pCi/1. The MDC and RL for Cs-134 in water are 15 and 30 pCi/I, respecti ve Iy. During preoperation, Cs-137 was detected in two of the samples at levels of 15 and 17 pCi/i. Then in 1984 and 1985, Cs-137 was again detected in a few samples with levels ranging from 4 to 5 pCi/i. The MDC and RL for Cs-137 in water are 18 and 50 pCi/I, respectively. Iodine-131 has never been detected in ground water samples. From 1986-2003, no radionuclides were detected. Since 2004, tritium has been detected at very low concentrations (near the instrument detection level) and close to environmental background concentration which is approximately 350 pCi/l (+/- 250 pCi/l) in the area around Farley. The positive results seen in these years were less than 3% of the reporting level for tritium. The MDC and RL for tritium in drinking water are 2,000 and 20,000 pCi/I, respectively. Figure 4.6-1 and Table 4.6-1 show the historical trending of the average annual detectable tritium concentration in offsite ground water. 4-27
Figure 4.6-1 Average Annual H-3 Concentration in Offsite Ground Water 2500 2000 - ~ 1500 . c: o ~ 1000 c: Q) CJ c: o o 500 K :\
~
)
~~ ~
\ 1\
........
o
.~ Ijl\V V I I) ~
Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year
-Indicator - Control - - MOC ]
4-28
Table 4.6-1 Average Annual H-3 Concentration in Offsite Ground Water Period Indicator Control (pCill) (pCiJI) Pre-op 150 240 1977 NDM NDM 1978 NDM 240 1979 NDM NDM 1980 124 NDM 1981 264 NDM 1982 240 NDM 1983 360 341 1984 NDM NDM 1985 NDM NDM 1986 NDM NDM 1987 NDM NDM 1988 NDM NDM 1989 NDM NDM 1990 NDM NDM 1991 NDM NDM 1992 NDM NDM 1993 NDM NDM 1994 NDM NDM 1995 NDM NDM 1996 NDM NDM 1997 NDM NDM 1998 NDM NDM 1999 NDM NDM 2000 NDM NDM 2001 NDM NDM 2002 NDM NDM 2003 NDM NDM 2004 194 271 2005 264 360 2006 NDM NDM 2007 218 321 2008 196 237 2009 474 401 2010 400 556 2011 NDM 333 2012 NDM NDM 4-29
As nuclear plants began to undergo decommissioning in the late 1990s to early 2000s, instances of subsurface and/or groundwater contamination were identified. In addition, several operating facilities also identified groundwater contamination resulting from spills, leaks, and equipment failure. In one instance, low levels of licensed material were detected in a private well located on property adjacent to a nuclear power plant. In 2006, NEI (Nuclear Energy Institute) formed a task force to address monitoring onsite groundwater for radionuclides at nuclear facilities. A Groundwater Protection Initiative was developed which was adopted by all U.S. commercial operating nuclear plants. The NRC also formed a task force to study the groundwater issues and released Information Notice 2006-13, Ground-water Contamination due to Undetected Leakage of Radioactive Water, which summarized its review of radioactive contamination of ground water at multiple facilities as a result of undetected leakage from structures, systems, and components that contain or transport radioactive fluids. Licensees were instructed to review the information for applicability and to consider appropriate actions to avoid similar problems. The NEI task force felt it was prudent for the industry to update site hydrology information and to develop radiological groundwater monitoring plans at each site. These groundwater protection plans would ensure that underground leaks and spills would be addressed promptly. Additionally, the task force recommended developing a communications protocol to report radioactive leaks or spills that have or could enter groundwater to the NRC and State / Local government officials as needed. NEI-07-07, Industry Groundwater Protection Final Guidance Document, was developed by the task force to document the guidelines recommended for the industry. To ensure compliance with NEI-07-07, Southern Nuclear developed the Nuclear Management Procedure, Radiological Groundwater Protection Program. The procedure contains detailed site-specific monitoring plans, program technical bases, and communications protocol (to ensure that radioactive leaks and spills are addressed and communicated appropriately). In an effort to prevent future leaks of radioactive material to groundwater, SNOC plants have also established robust buried piping and tanks inspection programs. In 2006, Farley located several old onsite piezometer wells and sampled these and the onsite drinking water wells for tritium and gamma isotopic activity. None of these wells contained detectable amounts of radioactivity. In 2007, after the site hydrology was evaluated, Farley implemented a more extensive radiological groundwater monitoring program which included drilling twelve new onsite monitoring wells (see Table 2-2). The twelve new wells along with one of the existing piezometer wells, the onsite drinking water wells, and several surface water / discharge locations comprise the monitoring program. These locations were sampled twice in the latter portion of2007 and sampled quarterly in 2008. Of the numerous samples taken from 2007 through 2012 (from the locations described above), only one location (groundwater well R-3) has shown low levels of radiological contamination (see Figure 4.6-2). 4-30
Tritium was the only nuclide identified. R-3 was also analyzed for gamma emitters (quarterly) and strontium (initially and after an increase was noted) and these were not detected. This well is located near the Protected Area and in the vicinity of the site where the Unit-2 radioactive effluent discharge line ruptured in 2002. In 2010, an Administrative Control Limit (ACL) was established for the area near R-3 where legacy material has been the source of tritium. The quarterly results for R-3 were all below the ACL of 6800 pCi/L of tritium . The ACL was derived based on previous years' tritium results and total measurement uncertainty. There are no reporting requirements associated with exceeding an ACL but additional actions would be taken to verify no new sources of tritium if an ACL was exceeded. In 2012, R-3 continued to be the only monitoring well with tritium concentrations consistently above the environmental background. The November 28, 2012 sample ofR-8 was slightly above the environmental background at 573 p/Ci/l (+/- 213 pCi/I). This well will be further evaluated if future results remain above background. Figure 4.6-2 H-3 Concentration in Onsite Ground Water Well R-3 7000 , 6000
/ :\
V'" 5000 L
~ .... .1
- s ~ 4000 ,- V- \ "- -o
/ I ~V ~ ~
c: - V-
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.-I -I-U 1000
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[ ,_ - - - R --=--=-- Admin Control L~ 4-3\
4.7 Surface Water As specified in Table 2-1 and shown in Figure 2-2, water samples are collected from the Chattahoochee River at a control station approximately 3 miles upstream of the intake structure and at an indicator station approximately 4 miles downstream of the discharge structure. Small quantities are collected during the week at periodic intervals using automatic samplers. For each station, one liter from each of four consecutive weekly samples is combined into a composite sample which is analyzed for gamma emitters. In addition, 0.075 liters is collected from 13 consecutive weekly samples for each station to fonn composite quarterly samples which are analyzed for tritium. No detectable results have been found from these gamma isotopic analyses since 1988. During preoperation and in every year of operation through 1988 (except 1979 and 1980), a few samples showed at least one of nine different activation or fission products at levels less than or on the order of their MDCs. During preoperation, Cs-137 was found in about 3% of the samples. From 1981 through 1988, it was found in about 15% of the samples. Cs-134 was found in about 15% of the samples from 1981 to 1986. All of these gamma emitters are attributed to the weapons tests and the Chemobyl incident. In 2012, as shown in Table 3-1, tritium was detected in 3 of the 4 quarterly composites at the indicator station (average 390 pCiIL) and in 1 of the 4 quarterly composite samples collected at the control station. All positive detections at both stations were in the range of the environmental background of approximately 350 pCi/L (+1- 250 pCi/L). Historical trending of the detectable concentrations of tritium in surface water is provided in Figure 4.7-1 and Table 4.7-l. The slightly elevated plot of the indicator stations is indicative of plant tritium contributions to surface water from pennitted plant effluent releases. However, it is noteworthy that the annual average levels are less than 10% of the MDC and less than 1% of the RL. The MDC and RL for tritium in surface water are 3,000 and 30,000, respectively. 4-32
Figure 4.7-1 Average Annual H-3 Concentration in Surface Water 3500 3000 2500 2 o ~2000 c: o ~ 1500 -g cQ): 1000 '--l - o U 500 o
~ b4 ~~
V ~ ~V ~~~
/' :---,
~
'-.., ~
~
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./
./~
~
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6 V I\. 8
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10 0--
/
12 Year
-Indicator - Control - MOC]
4-33
Table 4.7-1 Average Annual H-3 Concentration in Surface Water Period Indicator Control (~CiIl) (~Cill) Pre-op 200 170 1977 300 160 1978 230 250 1979 169 135 1980 221 206 1981 294 162 1982 300 132 1983 434 III 1984 333 152 1985 351 105 1986 478 272 1987 291.8 116.5 1988 293.3 NDM 1989 253 .8 NDM 1990 166 NDM 1991 122 NDM 1992 360.5 134 1993 388.8 NDM 1994 NDM NDM 1995 257 NDM 1996 386 NDM 1997 NDM NDM 1998 415 NDM 1999 314 NDM 2000 424 212 2001 252 NDM 2002 598 NDM 2003 296 NDM 2004 270 NDM 2005 215 173 2006 348 179 2007 321 NDM 2008 644 NDM 2009 343 NDM 2010 518 446 2011 401 NDM 2012 390 223 4-34
4.8 Fish Two types of fish (bottom feeding and game) are collected semiannually from the Chattahoochee River at a control station several miles upstream of the plant intake structure and at an indicator station a few miles downstream of the plant discharge structure. These locations are shown in Figure 2-2. Gamma isotopic analysis is performed on the edible portions of each sample as specified in Table 2-1. As shown in Table 3-1, Cs-13 7 was the only radionuclide of interest that was found from the gamma isotopic analysis of fish samples in 20 12. Cesium-137 was detected in both the spring and fall collection of game fish samples at the indicator station (average of 15.4 pCi/kg wet). Cesium-137 was not detected in either game fish sample at the control station. The MDC for Cs-13 7 in fish is 150 pCi/kg wet and the RL is 2000 pCi/kg wet. Cesium-137 was not detected in bottom feeding fish samples in either the spring or the fall sample collection at both the control and indicator locations. The MDC for Cs-13 7 in fish is 150 pCi/kg wet and the RL is 2000 pCi/kg wet. Figures 4.8-1 and 4.8-2 and Tables 4.8-1 and 4.8-2 provide the historical trending of the average annual detectable concentrations of Cs-137 in pCi/kg wet in bottom feeding and game fish, respectively. Since the early 1980s, values have generally decreased for both indicator and control groups, with the exception of the bottom feeding fish collected at the indicator station in 1993. While some contribution from the plant cannot be ruled out, most of the Cs-137 in these samples may be attributed to the nuclear weapons tests and the Chemobyl incident, as evidenced by the normally close agreement between the control and indicator station results. 4-35
Figure 4.8-1 Average Annual Cs-137 Concentration in Bottom Feeding Fish 250 200 ...g 100 L?-' - I~ C\2 t: Q) CJ 50 n
\/
t: o (J o Po 78 80
\V1\I 82 84 t:-. H 86 1\
88 90 92
/r\ [7' H "-Vl\ /
94 96 98 0
~
2 4 I 6
/'
8 lL ~ 10
- ",
12 Year
-+-Indlcator - Control - - MOC 4-36
Table 4.8-1 Average Annual Cs-137 Concentration in Bottom Feeding Fish Period Indicator Control (pCilkg) wet (pCi/kg) wet Pre-op 69 48 1977 NDM NOM 1978 NDM NDM 1979 38 30 1980 92 90 1981 96 106 1982 51.5 39.0 1983 NOM NOM 1984 NOM 19 1985 NOM NOM 1986 28 25 1987 25 19 1988 25.5 22.0 1989 NOM NOM 1990 NDM NOM 1991 NOM NDM 1992 NOM NDM 1993 208 NOM 1994 15.9 10.3 1995 NOM 14.2 1996 16.4 9.9 1997 10.9 7.7 1998 NOM NDM 1999 19.2 NDM 2000 NOM NOM 2001 9.8 NOM 2002 NOM NOM 2003 NOM 8.5 2004 8.1 NOM 2005 NOM 9.6 2006 9.7 NOM 2007 8.1 NOM 2008 11.4 7.7 2009 8.4 2l.9 2010 8.5 7.1 2011 10.0 4.3 2012 NOM NOM 4-37
Figure 4.8-2 Average Annual Cs-137 Concentration in Game Fish 350
-Q) ~
300 250 -Cl
~
( 3200 - Co -- c:
.S! 150 1 ~
c: B 100 v \\
~
c: o () 50 .~
-' u "- II 1\V 1\
~f I ~v ~ v' ~ ~
t-<
\ / 1\V i"--.
--V '"I'--.
'\.
V
'"- ~
. v, -.,., ./.
~ r-:--...
~
o Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year
-Indicator - Control - MD~
4-38
Table 4.8-2 Average Annual Cs-137 Concentration in Game Fish Period Indicator Control (pCilk~) wet (pCi/k~) wet Pre-op 84 60 1977 95 48 1978 NDM NDM 1979 111 83.5 1980 289 316 1981 189 126 1982 76 77 1983 57 56.5 1984 42 26 1985 84 44 1986 51 35 1987 83 46 1988 42 33 1989 38 29 1990 28 NDM 1991 36 24 1992 32.5 28 1993 34 NDM 1994 19 16 1995 17.9 18.2 1996 19.6 23.1 1997 25.9 NDM 1998 52 20 1999 36.9 15.9 2000 22.9 12.5 2001 22.4 12.3 2002 NDM 10.1 2003 19.3 12.0 2004 12.7 10.8 2005 15.7 NDM 2006 15.0 14.7 2007 15.4 6.5 2008 16.6 23.2 2009 24.9 12.5 2010 7.6 9.8 2011 9.0 15 .9 2012 15.4 NDM 4-39
Radionuclides of interest other than Cs-137 have been found in only a few samples in the past. The following table provides a summary of the results in pCi/kg wet compared with the applicable MDCs. YEAR Nuclide Fish Type Indicator Control MDC (pCilkg) (pCilkg) (pCilkg) 1978 Ce-144 Bottom Feeding NDM 200 1981 Nb-95 Bottom Feeding 38 NDM 50 (a) 1982 Nb-95 Game 31 NOM 50 (a) 1986 Co-60 Game 25 NDM 130 (a) Detennined by the EL. Not defined in ODCM Table 4-3 (Table 4-1 of this report) 4-40
4.9 Sediment River sediment samples are collected semiannually on the Chattahoochee River at a control station which is approximately 4 miles upstream of the intake structure and at an indicator station which is approximately 2 miles downstream of the discharge structure as shown in Figure 2-2. A gamma isotopic analysis is perfonned on each sample as specified in Table 2-1. During 2012, no man-made radioisotopes (nor Be-7) were detected in sediment collections. Although naturally occurring, Be-7 was also identified in the liquid effluents at Farley in 2012 . Historically, Be-7, Cs-134, Cs-137, and Nb-95 have been detected in some sampJes. These positive results were generally for samples collected at the control station. A summary of the positive historical results for these nuclides along with their applicable MOCs in units of pCilkg dry is provided in Table 4.9. Cs-134 and Cs-137 data are plotted in Figures 4.9-1 and 4.9-2, respectively. Table 4.9 Sediment Nuclide Concentrations Nuclide YEAR Indicator (pCilkg) Control (~Cilkg) MDC (pCilkgl Be-7 1985 535 945 655 (a) 2003 199 NOM 2009 72.8 NOM 2011 58.4 63.1 Cs-134 1987 NOM 45 150 1989 NOM 48 1992 138 51 1993 94 105 Cs-137 1981 NOM 185 180 1985 NOM 97 1989 NDM 39 1994 29 11 1996 11.8 NDM 2005 14.5 NOM 2009 NDM 24.4 2011 NDM 8.1 Nb-95 1981 52 113 50 (a) (a) Detennined by the EL. Not defined in OOCM Table 4-3 (TabJe 4-1 of this report). 4-41
Figure 4.9-1 Average Annual Cs-134 Concentration in Sediment 160
,
140
\
- Ci 120 ~ (3
\
-a. 100 ......... 0 IV c: 80 I I' c: Q) 60 U c: 0 (J 40 20 I \ I\ 1/
/ \11 \
o Po 78 80 82 84 86 88 90 92 94 96 98 0 2 4 6 8 10 12 Year
-+-Indlcator - Control The positive results for Cs-134 from 1986-1994 appear mostly at the control station. Due to its relatively short half-life of approximately 2 years, the positive results may be attributed to the Chemobyl incident. The overall plotting of the positive results does not show any discernible trends.
4-42
Figure 4.9-2 Average Annual Cs-137 Concentration in Sediment 200 180 I I -Cl ~ 140 160 () ~120 - r::: .2 100 -
~
CQ) 80 u g 60 l () 40 I \\ - 20 L\ 1\ o
/ \ ~ ~V ~ V l\ '/ \ /. ~
2 4 6 8 10 12 Year Cs-137 trended down after above ground weapons testing was stopped and has remained fairly low with random detections occurring. The majority of the positive results over the years are at the control station. Therefore in general, the positive results can be attributed to the weapons tests and the Chemobyl incident. 4-43
5.0 INTERLABORATORY COMPARISON PROGRAM In accordance with ODCM 4.1.3, the EL participates in an ICP that 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 the ICP is to be conducted with the Environmental Protection Agency (EPA) Environmental Radioactivity Laboratory Intercomparison Studies (Cross-check) Program or an equivalent program, and the ICP should include all of the determinations (sample mediumlradionuclide 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 Quality Assurance (QA) program and the capability to prepare Quality Control (QC) 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 perfonns the laboratory analyses in a normal manner. Each of the specified analyses is performed three times. The results are then sent to Analytics who perfonns 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 and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses of water samples. 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 as reported by Analytics. The uncertainty of the reported average is the propagated error of the values in the reported average by the EL. The precision of each result is measured by the coefficient of variation, which is defined as the standard deviation of the reported result 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: 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 umts are pCI/filter. For all other media, concentration units are pCi/liter (pCi/I).
5-1
As required by aDCM 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 and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses 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. The Environmental Radiochemistry laboratory participates in a performance evaluation (PE) sample program provided by Analytics Inc. The PE samples are received and analyzed routinely with environmental and effluent samples. The laboratory analyzed 9 samples for 35 parameters in 2012. The 2012 analyses included tritium, gross beta and gamma emitting radio-nuclides in different matrices. The attached results indicate 3 analyses (Ce-141, Cr-51, and Fe-59) were outside the acceptance limits for accuracy. These isotopes were in the Gamma in Air Filter matrix. After the results were received, the sample was recounted but two of the isotopes had decayed off. The remaining isotopes were within acceptable limits for accuracy. A Gamma in Air Filter PE sample will be analyzed in 2Q 2013 to complete an investigation. 5-2
TABLE 5-1 (SHEET 1 of3) INTERLABORATORY COMPARISON PROGRAM RESULTS 1-131 ANALYSIS OF AN AIR CARTRIDGE (pCilcartridge) Analysis or I Date I Reported JKnown Standard Uncertainty Percent Coef I Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation 1-131 1 07/14112 1 100.00 1 97.20 6.45 1.62 7.221 0.44 GAMMA ISOTOPIC ANALYSIS OF AN AIR FILTER (pCilfIlter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Anal)'tics !3Sj of Variation Deviation Ce-141 09/13112 113.00 153.00 6.66 2.56 7.72 0.02 Co-58 09/13/12 77.50 94.10 3.85 1.57 7.41 -2.90 Co-60 09113 / 12 117.00 142.00 4.14 2.38 5.69 -3.82 Cr-51 09113112 184.00 232.00 18.3 3.88 14.27 -1.84 Cs-134 09113112 83.40 101.00 2.73 1.69 5.39 -3.92 Cs-137 09113112 130.00 163.00 4.83 2.73 5.96 -4.22 Fe-59 09113112 111.00 142.00 7.18 2.38 8.76 -3.14 Mn-54 09/13112 150.00 183 .00 12.5 3.06 9.49 -2.29 Zn-65 09113112 154.00 180.00 13.4 3.01 10.16 -2.29 GROSS BET A ANALYSIS OF AN AIR FILTER (pCilfiUer) Analysis or Date Reported I Known I Standard . Uncertainty Percent Coef I Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 09/13112 93.00 1 84.10 1 1.2 1.40 4.21 1 2.27
TABLE 5-1 (SHEET 2 of3) INTERLABORA TORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF A MILK SAMPLE (pCilliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Ce-141 7114112 76.60 82.20 2.85 1.37 8.62 -0.84 Co-58 7114112 90.20 99.70 2.85 1.66 7.33 -0.32 Co-60 7114112 350.00 355.00 10.7 5.93 4.23 -0.35 Cr-51 7114112 433 .00 402.00 21.1 6.71 8.51 0.85 Cs-134 7114112 180.00 174.00 6.26 2.91 4.68 0.69 Cs-137 7114112 216.00 212.00 9.26 3.54 5.63 0.34 Fe-59 7114/ 12 126.00 128.00 6.53 2.13 8.21 -0.17 1-131 7114112 102.00 99.70 7 1.66 8.78 0.30 Mn-54 7114112 134.00 132.00 3.79 2.21 5.69 0.24 Zn-65 7114112 208.00 199.00 8.17 3.33 7.09 0.59 GROSS BET A ANALYSIS OF WATER SAMPLE (pCilliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 03115 /12 263 .00 297.00 18.93 4.96 1.10 -11.99 0711 4112 166.00 148.00 10.28 2.47 10.85 0.98 GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES (pCilliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Avera2e Value Deviation EL Analytics (3~ of Variation Deviation Ce-141 03115112 190.00 184.00 5.65 3.07 5.25 0.64 Co-58 03115112 92.50 93.40 5.19 1.56 8.20 -0.11 Co-60 03115112 208.00 197.00 5.68 3.29 4.59 1.16
TABLE 5-1 (SHEET 3 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES CONT. (pCilliter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Cr-51 03/15112 362.00 309.00 46.7 5.16 15.63 0.94 Cs-134 03115112 108.00 106.00 l.75 l.77 4.41 0.49 Cs-137 03115112 124.00 113.00 3.67 l.88 6.09 1.50 Fe-59 03115/12 119.00 119.00 6.14 1.99 8.23 0.02 I-131 03115112 104.00 93.80 4.15 1.57 6.68 1.44 Mn-54 03115/12 149.00 138.00 2.24 2.31 5.13 1.38 Zn-65 03/15112 245.00 235.00 5.41 3.93 5.98 0.67 TRITIUM ANALYSIS OF WATER SAMPLES (pCilliter) VI I VI Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation H-3 03115/12 4160 4470 102.54 74.70 4.43 -1.70 07114112 4580 4970 92.36 83.00 4.34 -1.98
6.0 CONCLUSION
S This report confirms the licensee's conformance with the requirements of Chapter 4 of the ODCM. It provides a summary and discussion of the results of the laboratory analyses for each type of sample. In 2012, there was one instance in which the indicator stations results were statistically discernible from the control station results. Cesium-137 was also detected in game fish samples in both the spring and fall sample collections at the indicator location, FGI-S5. The average Cs-137 of the two samples was 15.4 (pCi/kg) wet. Using the ingestion dose factors and consumption rate factors in Reg. Guide 1.109 it was calculated that the highest potential dose to a maximum exposed member of the public (an adult), due to regular consumption of fish containing Cs-137 at the low level seen in the 2012 samples, would be approximately 2.31E-2 mrem in a year. This dose is about 0.8% of the regulatory limit of 3 mrem per year due to liquid effluents. While the Cs-137 seen in the game fish could be attributed to plant effluents, low levels of Cs-137 in the environment are attributed primarily to fallout from nuclear weapons testing and from the Chernobyl incident. 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 other REMP samples. The radiological levels reported from 2012 sample results remained low. The REMP trends over the course of time from preoperation to the present are generally decreasing or have remained fairly constant at low levels. This supports the conclusion that there is no adverse radiological impact on the environment or to the public as a result of the operation of Farley Nuclear Plant. 6-1
7.0 ERRATA The following pages are corrections to the Joseph M. Farley Nuclear Plant Annual Radiological Environmental Operating Report for 2011 . The corrections are a result of the discovery, by Georgia Power Company Environmental Laboratory staff in 2012, of a small positive bias in the 2011 results of OSL environmental dosimeter readings. The method used during 2011 was acceptable at the time but EL dosimetry personnel studied the source of the bias and determined it was based on higher residual dose on the OSL badges as compared to the past Panasonic system. New processing methods are now in place and included in processing procedures. All 2012 environmental OSL processing and reports have included the new methods to remove this small positive bias. The correction has been applied to the 2011 OSL dosimeter results and the corrected data are described in the following pages. 7-1
TABLE 3-1 (SHEET 1 of 6) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
Farley Nuclear Plant, Docket Nos. 50-348 and 50-364 Houston County, Alabama Medium or I Type and Minimum Indicator Location with the Highest Community Control Pathway Sampled Total Detectable Locations Annual Mean Locations Locations (Unit of Number of Concentration Mean (b), Mean (b), Mean(b), Measurement) Analyses (MDC) (a) Range N arne Distance Mean (b), Range Range Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 20.9 PC-1101 19.4 18.9 17.3 Particulates 459 2.8-43.4 Plant Entrance 5.6-37.1 4.4-43.3 3.2-50.3 (fCi/m3) (205/205) 0.9 miles (49/49) (153/153) (104/ 104) WSW
-_..... - -_. __.._ .. __.........*... .... _._-- ..._-----_ .._- --_.--------
,
---------------...--
Gamma Isotopic 36 Be-7 24 73.2 PI-0701 83.2 68.5 72.0 47 .0-109.5 S. Perimeter 66.1-109.5 53.0-89.1 55.0-100.9 (16/16) 1.0 miles SSE (4/4) (12/12) (8/8) -.l W 1-131 70 NDM NA NDM NDM NDM N I I N (0/16) (0/16) (0/12) (0/8) Cs-134 50 NDM(c) NA(d) NDM NDM (0/16) (0/ 12) (0/8) Cs-137 60 NDM NA NDM NDM (0/16) (0/12) (0/8) Airborne 1-131 70 NDM NA NDM NDM NDM Radioiodine 359 (01203) (0/52) (0/104) ifCi/m3) Direct Radiation Gamma NA 15.1 Rl-0401 23.8 13.0 14.4 (mRl91 days) Dose 1.1-25.0 Pint Perimeter 22.3-25.0 9.0-16.3 8.8-17.9 156 (64/64) 0.8 miles E (4/4) (72172) (24124) Milk (PCi/l) Gamma Isotopic 0 Cs-134 15 NA NA NA NA Cs-137 18 NA NA NA NA Ba-140 60 NA NA NA NA La-140 1-131 15 1 NA NA
-
NA p * * - . _ _ ** * * - NA
** , *
- _ - - - _ ** _ _ . _ - - _ ** * ' . , . _ *** -_._-- _.......... _.*._.
NA NA NA NA---------- -
4.3 Direct Radiation In 2011, direct (external) radiation was measured with Landauer InLight optically stimulated luminescent (OSL) dosimeters which replaced the Panasonic thennoluminescent dosimeters (TLDs). The Panasonic system was retired at the end of 20 10 due to the inability to keep the aging badge readers operating reliably. Similar to the TLD protocol of the past, two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 perfonned near mid-quarter for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges. Direct radiation stations are established in each of the 16 sectors, to fonn 2 concentric rings. The inner ring stations are located near the plant perimeter, as shown in Figure 2-1, and the outer ring stations are located at distances of approximately 3 to 5 miles from the plant, as shown in Figure 2-2. The stations fonning the inner ring are designated as the indicator stations. The 6 control stations are located at distances greater than 10 miles from the plant, as shown in Figure 2-3. Stations are also provided which monitor special interest areas: the nearest occupied residence (SWat 1.2 miles), as shown in Figure 2-1, and the city of Ashford (WSW at 8 miles), as shown in Figure 2-3. The 16 outer ring stations and the 2 special interest stations are designated as community stations. As provided in Table 3-1, the average quarterly exposure measured at the indicator stations (inner ring) during 2011 was 15.1 mR which was 0.7 mR greater than the 14.4 mR which was acquired at the control stations. This difference is not statistically discernible since it is less than the MDD of 2.0 mR. The difference of 1.4 mR found between the control stations (14.4 mR) and community stations (13 .0 mR) is statistically discernible since the difference is slightly greater than the MDD of 1.2 mR. The discernible difference does not indicate that Plant Farley effluents are affecting the environment because the community station readings are actually lower than the control stations. The difference between the indicator and control and between the control and community stations is consistent with what has been seen in previous years. The historical trending of the average quarterly exposures in units of mR at the indicator, control, and community locations are plotted in Figure 4.3 -1 and listed in Table 4.3-1. During preoperation the average exposure at the indicator stations was 1.2 mR greater than that for the control stations, but the average over the entire period of operation was only 1.1 mR greater. During preoperation, the average exposure at the control stations was 1.3 mR greater than that at the community stations and the average over the period of operation was 1.5 mR greater. This supports the position that the plant is not contributing significantly to direct radiation in the environment. 4-14 7-3
In 2011, only the OSL badges were placed on station. During 2010, the OSL badges had been co-located on station with the TLD badges. Following the change to only OSL badges, the differences between indicator, control, and community locations is consistent with previous years. An increase noted in 2010 nd reflects issues (especially during 2 Qtr) with the aging Panasonic TLD reader. Table 4-3 lists the REMP program deviations that occurred in 2011. There were no deviations involving OSL badges. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 30 25 -20 ~l jl\ ~D.. JI ~ ~ ~ V~ ~ ~v 0:: 1/ (J k- ~ - E
~ 15
~
~If
~ iF""
~ VJ
~
'Iif I" r--, ~ ~ V V f'.,
~ !--.. ['-,
v:. ,""..-. V) ~ III 1\ V t) ~
- J In 'f' "l:Z"" r o
0.10
)(
1 W 5 o
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Year
-Indicator ___ Control ---.- Community 4-15 7-4
Table 4.3-1 Average Quarterly Exposure from Direct Radiation Period Indicator (mR) Control (mR) Community (mR) Pre-op 12.6 11.4 10.1 1977 10.6 12.2 10.6 1978 15.0 13.5 12.0 1979 20.3 18.7 15.2 1980 21.9 21.6 18.5 1981 16.5 14.9 14.5 1982 15.5 14.7 13.0 1983 20.2 20.2 17.4 1984 18.3 16.9 15.3 1985 21.9 22.0 18.0 1986 17.8 l7.7 15.1 1987 20.8 20.0 18.0 1988 21.5 19.9 18.5 1989 18.0 16.2 15.3 1990 18.9 16.4 15.8 1991 18.4 16.1 16.1 1992 16.1 13.6 13.5 1993 17.4 15.9 15.6 1994 15.0 13.0 12.0 1995 14.0 12.5 11.8 1996 14.2 12.7 11.9 1997 15.3 13.9 11.9 1998 16.2 14.6 13.9 1999 14.7 13.4 12.6 2000 15.5 14.1 13.5 2001 14.9 13.4 12.7 2002 14.1 12.6 11.9 2003 15.2 13.6 12.9 2004 14.3 12.9 12.1 2005 14.7 13.4 12.5 2006 15.2 13.6 12.9 2007 14.6 13.3 12.5 2008 15.0 13.7 12.9 2009 15.2 13.6 12.8 2010 17.8 16.7 15.5 2011 15.1 14.4 13.0 4-16 7-5
The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self imposed limit of 3.S. Previously with TLDs, this limit had been 1.4. However, the OSL readings varied more (between the two elements) than the TLD readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication lSD, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 2010. This limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater than or equal to 3.S are excluded since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 2011, the OSL results from the following stations were excluded from the data set because their standard deviations were greater than 3.S: Quarter 1 - Rl-0301B Quarter 2 - RI-0101A, Rl-0301A, Rl-0601B, RI-1S01B, Rl-1601A, RC-OSOSB, and RC-11 04A Quarter 3 - None Quarter 4 - None For the direct radiation stations where these badges were located, only the reading of the companion badge was used to determine the quarterly exposure for the station. The badges (with::: 3.S SD) 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 found for the high standard deviations. A major advantage of the OSL badge is that it can be read multiple times. A new practice was employed in 2011 to re-read any environmental badges that yielded a standard deviation::: 3.S. The readings with the lower standard deviation would be reported. 4-17 7-6
6.0 CONCLUSION
S This report confirms the licensee's conformance with the requirements of Chapter 4 of the ODCM. It provides a summary and discussion of the results of the laboratory analyses for each type of sample. In 2011, there were three instances where the indicator stations results were statistically discernible from the control station results. The annual average weekly gross beta activity in air filters at the indicator stations was statistically discernible from the control stations, and community stations were also statistically discernible from the control station results. The annual average weekly gross beta activity in air filters was 20.9 fCilm 3 at the indicator stations and 14.5 fCi/m 3 at the control stations. The average at the community stations was 18.2 fCi/m 3. Gross beta is a screening ana'1sis for beta activity. The required MDC for gross beta in air filters is 10 fCi/m ; there is no Reporting Level for gross beta in air. 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's contribution to gross beta concentration in air is insignificant. In the fall 2011 fish collection, Cs-137 activity in bottom-feeding fish was slightly higher at the indicator station (10 pCi/kg-wet) than at the control station (4.3 pCi/kg-wet). No Cs-137 was detected in the spring collection for bottom-feeding fish. Cesium-137 was detected in both the fall and spring collection (at both the indicator and control stations) for game fish but there was no discernible difference in the results. The potential dose to a member of the public who would receive the highest dose (an adult) due to regular consumption of fish containing Cs-137 at the low level seen in the fall would be approximately 8.55E-3 mrem in a year. This dose is about 0.3% of the regulatory limit of 3 mrem per year due to liquid effluents. While the Cs-137 seen in bottom-feeding fish could be attributed to plant effluents, low levels of Cs-137 in the environment are attributed primarily to fallout from nuclear weapons testing and from the Chernobyl incident. There was also a statistically discernible difference between the control and community direct radiation stations. Because the community stations (13.0 mR) indicated less than the controls (14.4 mR), this discernible difference does not indicate that plant effluents are affecting the environment. 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 other REMP samples. The radiological levels reported in 2011 were low and are generally trending downward. The REMP trends over the course of time from preoperation to the present are generally decreasing or have remained fairly constant. This supports the conclusion that there is no adverse radiological impact on the environment or to the public as a result of the operation of Farley Nuclear Plant. 6-1 7-7
Edwin I. Hatch Nuclear Plant Joseph M. Farley Nuclear Plant Vogtle Electric Generating Plant Annual Radiological Environmental Operating Reports for 2012 Enclosure 3 Vogtle Annual Radiological Environmental Operating Reports for 2012
VOGTLE ELECTRIC GENERATING PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 SOUTHERN COMPANY Energy to Serve Your World'"
VOGTLE ELECTRIC GENERATING PLANT ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT FOR 2012 April 26, 2012 FINAL hen1. Tim Meents Ian Lake Tim .Meents@chemstaff.com lan.Lake@chemstaff.com 815-600-9247 815-600-2067 Dennis Oltmans DOltmans@chemstaff.com 717-575-3481
TABLE OF CONTENTS Section and/or Title Subsection Pa~e List of Figures II
...
List of Tables 111 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 4-5 4.2 Airborne 4-7 4.3 Direct Radiation 4-11 4.4 Milk 4-17 4.5 Vegetation 4-19 4.6 River Water 4-23 4.7 Drinking Water 4-26 4.8 Fish 4-36 4.9 Sediment 4-39
- 4. 10 Groundwater 4-48 5.0 Interlaboratory Comparison Program (ICF') 5-1 6.0 Conclusions 6-1 7.0 Errata 7-1
LIST OF FIGURES Figure Number Title Page Figure 2-1 REMP Stations in the Plant Vicinity 2-11 Figure 2-2 REMP Control Stations for the Plant 2-12 Figure 2-3 REMP Indicator Drinking Water Stations 2-13 Figure 2-4 Groundwater Monitoring Wells 2-14 Figure 4.2-1 A verage Weekly Gross Beta A ir Concentration 4-8 Figure 4.3-1 A verage Quarterly Exposure from Direct Radiation 4-12 Figure 4.3-2 A verage Quarterly Exposure from Direct Radiation at Special Interest Areas 4-14 Figure 4.4-1 Average Annual Cs-137 Concentration in Milk 4-17 Figure 4.5-1 A verage Annual Cs-137 Concentration in Vegetation 4-21 Figure 4.6-1 Average Annual H-3 Concentration in River Water 4-24 Figure 4.7-1 A verage Monthly Gross Beta Concentration in Raw Drinking Water 4-27 Figure 4.7-2 A verage Monthly Gross Beta Concentration in Finished Drinking Water 4-29 Figure 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water 4-32 Figure 4.7-4 Average AIIDual H-3 Concentration in Finished Drinking Water 4-34 Figllre 4.8-1 A verage Annual Cs-137 Concentration in Fish 4-37 Figure 4.9-1 A verage Annual Be-7 Concentration in Sediment 4-40 Figure 4.9-2 Avera~e Annual Co-58 Concentration in Sediment 4-42 Figure 4.9-3 A verage Annual Co-60 Concentration in Sediment 4-44 Figure 4.9-4 Average Annual Cs-137 Concentration in Sediment 4-46 Figure 4.10-1 Ground Water H3 Concentration in Protected Area Water Table Wells 4-50 Figure 4.10-2 Ground Water H3 Concentration in Existing Water Table Wells 4-51 Figure 4 . 10-3 Ground Water H3 Concentration in New Water Table Wells 4-52 ii
LIST OF TABLES Table Number Title Pa2e Table 2-1 Summary Description of Radiological Environmental Monitoring Program 2-2 Table 2-2 Radiological Environmental Sampling Locations 2-7 Table 2-3 Groundwater Monitoring Locations 2-10 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 4-5 Table 4.2-1 A verage Weekly Gross Beta Air Concentration 4-9 Table 4.3-1 A vemge Quarterly Exposure from Direct Radiation 4-13 Table 4 .3-2 A verage Quarterly Exposure from Direct Radiation at Speciallnterest Areas 4-15 Table 4.4-1 A verage Annual Cs-137 Concentration in Milk 4-18 Table 4.5-1 A verage Annual Cs-13 7 Concentration in Vegetation 4-22 Table 4 .6-1 Average Annual H-3 Concentration in River Water 4-25 Table 4 .7-1 A verage Monthly Gross Beta Concentration in Raw Drinking Water 4-28 Table 4.7-2 A verage Monthly Gross Beta Concentration in Finished Drinking Water 4-30 Table 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water 4-33 Table 4 .7-4 Average Annual H-3 Concentration in Finished Drinking Water 4-35 Table 4.8-1 A verage Annual Cs-13 7 Concentration in Fish 4-38 Table 4.9-1 A verage Annual Be-7 Concentration in Sediment 4-41 Table 4.9-2 A verage Annual Co-58 Concentration in Sediment 4-43 Table 4.9-3 A verage Annual Co-60 Concentration in Sediment 4-45 Table 4.9-4 A verage Annual Cs-137 Concentration in Sediment 4-47 Table 4.9-5 Additional Sediment Nuclide Concentrations 4-47 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) NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual OSL Optically Stimulated Luminescence 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 2012 are reported herein in accordance with Technical Specification (TS) 5.6.2 and ODCM 7.1. The objectives of the REMP are to: I) 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), areas of higher population (community 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 3626 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 3626 MWt, received its operating license on February 9, 1989 and began commercial operation on May 19,1989. The pre-operational 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. I-I
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-I . 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-J through 2-3. Figure 2-4 indicates the locations of onsite groundwater wells. These wells are not part of the REMP but are part of the VEGP Radiological Groundwater Protection Program describe in section 4.10. REMP samples are collected by Georgia Power Company's (GPC) Environmental Laboratory (EL) personnel. The same lab perfolms all the laboratory analyses at their headquarters in Smyrna, Georgia . 2-1
TABLE 2-1 (SHEET 1 of5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Collection Type and Frequency of and/or Sample Samples and Sample Frequency Analysis Locations
- 1. Direct Radiation Thirty nine routine monitoring Quarterly Gamma dose, quarterly 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 Samples from seven locations: Continuous sampler operation Radioiodine canister: l-and Particulates with sample collection weekly, 131 analysis, weekly.
Five locations close to the site or more frequently if required by boundary in different sectors; dust loading. Particulate sampler: Gross beta analysis I A community having the following filter change highest calculated annual and gamma isotopic average ground level D/Q; analysis 2 of composite (by location), quarterly. A control location near a population center at a distance of about 14 miles.
TABLE 2-1 (SHEET 2 of 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Collection Type and Frequency of and/or Sample Samples and Sample Frequency Analysis Locations
- 3. Waterborne
-a~--S-u-[fac(?*---------------- ---OnesampTe-upnver.- - - *-- -- Composite sample over one --- oammaTsotoplc------ ----
4 month period . analysis 2 , monthly. Two samples downriver. Composite for tritium analysis, quarterly.
****5~-Dnnklng-- --- - ----**--***---*-* *--Two sam-ples-at**ea-cil-oTihe----*-*-***t----:C==-o-m-po-s--=i-te-s-a-m-p-'l'--e-o-'f;-r-=-iv-e-r-w-a-t-er---l-** T~T3Tana IYSIs-on-each---
three nearest water treatment near the intake of each water sample when the dose plants that could be affected treatment plant over two week calculated for the 4 by plant discharges. period when 1-131 analysis is consumption of the required for each sample; water is greater than 1 Two samples at a control monthly composite otherwise; mrem per year). location. and grab sample of finished Composite for gross water at each water treatment beta and gamma plant every two weeks or isotopic analysis 2 on N, w monthly, as appropriate. raw water, monthly. Gross beta, gamma isotopic and I-l31 analyses on grab sample of finished water, monthly. Composite for tritium analysis on raw and finished water, quarterly.
- c. Groundwater See Table 2-3 and Figure 2-4 Quarterly sample; pump used to Tritium, gamma isotopic, for well locations . sample GW wells; grab sample and field parameters (pH, from yard drains and ponds temperature, conductivity, dissolved oxygen, oxidation/reduction potential, and turbidity) of each sample quarterly; Hard to detect radionuclides as necessary based on results of tritium and gamma
TABLE 2-1 (SHEET 3 ofS)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Collection Type and Frequency of and/or Sample Samples and Sample Frequency Analysis Locations
- d. Sediment from Shoreline One sample from downriver Sem iannuall y Gamma isotopic area with existing or potential 2 analysis , semiannually.
recreational value. One sample from upriver area with existing or potential recreational value. -}*~f~s-~~~~....- -- . . ----.-- * *- _ * **-*- - Two-san1ples-from-mlTklng--*-*-*I--= B-o-im-o-n-:th-:l-y---------i. . --Gammalso-to.plc.-.---.... . --.. --. 6 animals at control locations analysis 2,7, bimonthly. at a distance of about 10 miles or more.
- -b.P is"fi- - *-*-**- ----- ------- At-Ieast-one-s-ample*ofany-- -* *-I----:::S=-e-m-:i-ann-u-a-:l.,-ly--------+- G~lmma-lsotopl- c ---**-******-*-**-
2 commercially or analysis on edible recreationally important portions, semiannually. 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 any During the spring spawning Gamma isotopic anadromous species near the season. analysis 2 on edible plant discharge. portions, annually.
TABLE 2-1 (SHEET 4 of 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Collection Type and Frequency of and/or Sample Samples and Sample Frequency Analysis Locations
- c. Grass or Leafy One sample from two onsite Monthly during growing Gamma isotopic Vegetation locations near the site season. analysis 2. 7, monthly.
boundary in different sectors. One sample from a control location at a distance of about 17 miles. N I VI
TABLE 2-1 (SHEET 5 of 5)
SUMMARY
DESCRIPTION OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Notes: (I) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours 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.
,
tv (6) A milking animal is a cow or goat producing milk for human consumption. 0\ (7) If the gamma isotopic analysis is not sensitive enough to meet the Minimum Detectable Concentration (MOC) for I-131 , a separate analysis for I-131 may be performed.
TABLE 2-2 (SHEET 1 of 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Station Station Descriptive Direction l 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.l3.1 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 Direction l Distance Sample Type Number Type Location (miles)) 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. HQ 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 Water 2 4 81 Control Sav River N 2.5 Fish 3 Sediment 82 Control Sav River (RM NNE 0.8 River Water 151. 21 83 Indicator Sav River (RM ENE 0.8 River Water 4 150.4) Sediment 84 Other Sav River (RM ESE 1.6 River Water 149.5) 85 Indicator Sav River ESE 4.3 Fish 3 87 Indicator Beaufort-Jasper SE 76 Drinking County Water WaterS Treatment Plant 88 Indicator Cherokee Hill SSE 72 Drinking Water Treatment Water6 Plant, Port Wentworth, Ga 2-8
TABLE 2-2 (SHEET 3 of 3) RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS 89 Indicator Purrysburg SSE 76 Drinking Water Treatment Water7 Plant; Purrysburg, SC 98 Control W.e. Dixon SE 9.8 Milk~ Dairy 101 Indicator Girard Dairy S 5.5 Milko 102 Control Seven Oaks W 7.5 Milko Dairy 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 unavai [able 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 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 Abercom Creek which is about one and a quarter creek miles from its mouth on the Savannah River at RM29. (7) The intake for the Purrysburg Water Treatment Plant is located on the same canal as the Beaufort-Jasper Water Treatment Plant. The Purrysburg intake is nearer to the Savannah River at the beginning of the canal. (8) Girard Dairy is considered an indicator station since it is the closest dairy to the plant (@5.5 miles). Dixon Dairy went out of business in June 2009 and Seven Oaks Dairy (@7.5 miles) was added as a replacement and is considered a control station even though a control station is typically 10 miles or greater. 2-9
Groundwater Monitoring Locations Tab1e 2-3 WELL AQUIFER MONITORlNG PURPOSE LT-IB Water Table NSCW related tank LT-7A Water Table NSCW related tank LT-12 Water Table NSCW related taru( LT-13 Water Table NSCW related tank 802A Water Table Southeastern potential leakage 803A Water Table Up gradient to rad waste building Down gradient from rad waste bldg and 805A *** Water Table NSCW related facilities 806B Water Table Dilution line 808 Water Table Up gradient; along Pen Branch Fault NSCW related tank; western potential Rl Water Table leakage R2 Water Table Southern potential leakage R3 Water Table Eastern potential leakage R4 Water Table Dilution line R5 Water Table Dilution line R6 Water Table Dilution line R7 Water Table Dilution line Water Table within R8 Dilution line Sav. River sediments 1013* Water Table Low level rad waste storage 1014 Tertiary Up gradient 1015 Water Table Vertically up gradient 1003* Tertiary Up gradient 1004* Water Table Vertically up gradient 27** Tertiary Down gradient tertiary 29** Tertiary Down gradient tertiary MU-l Tertiary/Cretaceous Facility water supply River N/A Surface water NSCW - Nuclear service cooling water
- Wells abandoned in Feb. 2009 due to construction activities with proposed new units
- Sampling discontinued in 2010 due to structural issues with the well
- Well abandoned in 2009 due to structural issues 2-10
- Sampling discontinued in 2010 due to structural issues with the well
56
~N Radiological Environmental Sampling Locations Indicator Control Additional REMP Stations TLD * *
- in the Plant Vicinity Other * *
- TLD & Other ;\
Figure 2-1 2-11
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3.0 RESULTS
SUMMARY
In accordance with aDCM 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 aDCM 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 radiolluclide Be-7, which occurs abundantly in nature, is often detected in REMP samples. It is occasionally detected in the plant ' s liquid and gaseous effluents. When it is detected in effluents, it is also included in the REMP results. In 2012, Be-7 was not detected in either liquid and gaseous effluents at Vogtle. 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 Other Control Pathway Total Detectable Locations Annual Mean Stations (g) Locations Sampled Number of Concentration Mean (b), Mean (b), Mean (b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Range Range Measurement) Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 25.9 Station 10 27.6 26.1 25.2 Particulates 371 4.5-51.4 Met Tower 13.4-49.3 12.3-48.37 8.6-50.0 (fCi/m3) (265/265) 0.9 miles SSW (53/53) (53/53) (53/53)
"- -"'- '-" --" '-'. ' -~ -""" -'- "'-'
Gamma
- .*.............. _-_.... __ _
......... ........... __ ..... - ..*..". .*.. _...*. __..*.....*...._-_......*......_- . .. _._.._.__ .._._..... _-_.._...._.. _- .......__ ._._-- .... _--..- ......*.__... _ -_ .....*...*.... -_ __ _ .
......* *... . __ _
..... ....-._-----.. __.. -
Isotopic 28 Be-7 24 87.0 Station 10 97.2 83.8 82.0 63.2-126.7 Met Tower 68.9-120 .9 67.0-102.2 57.0-109.2 (20/20) 0.9 miles SSW (4/4) (4/4) (4/4) 1-131 70 NOM NOM NOM NOM Cs-134 50 NOM (c) NOM NOM NOM Cs-137 60 NOM NOM NOM NOM Airborne 1-131 70 NOM NA NOM NOM Radioiodine 371 (fCi/m3) Direct Gamma NA (d) 14.4 Station 0 I 19.8 14.2 14.3 Radiation Dose 9.6-22.7 River Bank 19.1-20.4 9.4-20.2 11.6-17.3 (mRl91 days) 151 (64/64) 1.1 mile N (4/4) (64/64) (23/23)
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 Other Control Pathway Total Detectable Locations Annual Mean Stations (g) Locations Sampled Number of Concentration Mean (b), Mean (b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Mean (b), Range Measurement) Performed (Fraction) & Direction Range (Fraction) Range (Fraction) (Fraction) Milk (pCi/l) Gamma Isotopic 48 1-131 1 NOM NA NOM NA NOM Cs-134 15 NOM NOM NA NOM Cs-137 18 NOM NOM NA NOM w, Ba-140 60 NOM NOM NA NOM w La-140 15 NOM
....... ..... _--._ ...........
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1-131 1 NOM NA NOM NA NOM 48 Vegetation Gamma (pCi/kg-wet) Isotopic 36 Be-7 729 2724.6 Station 15 3846.1 NA 2063.8 274.1-13562 Hancock Landing Rd 711.2-15399 621.8-3870.9 (24/24) 1.5 miles NW (12/ 12) (12112) 1-131 60 NOM NA NOM NA NOM Cs-134 60 NOM NA NOM NA NOM Cs-137 80 NOM NA NOM NA NOM
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 Other Control Pathway Total Number Detectable Locations Annual Mean Stations (g) Locations Sampled of Analyses Concen tra tion Mean (b), Mean (b), (MOC) (a) Mean (b), (Unit of Performed Range Name Distance Mean (b), Range Measurement) (Fraction) & Direction Range (Fraction) Range (Fraction) (Fraction) River Water Gamma (pCi/i) Isotopic 36 8e-7 124(e) NOM NOM NOM NOM Mn-54 15 NOM NOM NOM NOM Fe-59 30 NOM NOM NOM NOM Co-58 15 NOM NOM NOM NOM Co-60 15 NOM NOM NOM NOM Zn-65 30 NOM NOM NOM NOM Zr-95 30 NOM NOM NOM NOM Nb-95 15 NOM NOM NOM NOM 1-131 15 NOM NOM NOM NOM Cs-134 15 NOM NOM NOM NOM Cs-137 18 NOM NOM NOM NOM 8a-140 60 NOM NOM NOM NOM La-140 15 _--_.. NOM ____ ..*... _-._ _-_ _. __...__ NOM NOM __.....* _--_....- .... __ .- _--_.__......*._..._... -_..... *........... ......- .....*.----- ..-.. .*.* --.--.-- -.-.---..---..--.--.*..-- -.. NOM Tritium 200Cy--_*** - _
.... .. ......
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412
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12 393-3310 RM 150.4 393-3310 258-636 (114) (4/4) 0.8 miles ENE (4/4) (4/4)
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 Total Minimum Indicator Location with the Highest Other Control Pathway Number of Detectable Locations Annual Mean Stations (g) Locations Sampled Analyses Concen tra tion Mean (b), Mean (b), Mean (b), (Unit of Performed (MDC) (a) Range Name Distance Mean (b), Range Measurement) (Fraction) & Direction Range (Fraction) Range (Fraction) (Fraction) Water Near Gross Beta 4 3.83 Station 87 3.85 NA 2.65 Intakes to 48 2.22-6.62 Beaufort-Jasper County 2.30-5.62 1.13-4.73 Water (36/36) Water Treatment Plant (12/ 12) (l21l2) Treatment 76 miles SE Plants (pCi/l) Gamma Isotopic w, 48 V> Be-7 124(e) NOM NOM NA NOM Mn-54 15 NOM NOM NA NOM Fe-59 30 NOM NOM NA NOM Co-58 15 NOM NOM NA NDM Co-60 15 NOM NOM NA NOM Zn-65 30 NOM NOM NA NOM Zr-95 30 NOM NOM NA NOM Nb-95 15 NOM NOM NA NOM 1-131 (f) 15 NDM NOM NA NOM Cs-134 15 NOM NOM NA NOM Cs-137 18 NOM NOM NA NOM Ba-140 60 NOM NOM NA NOM 15 NDM NOM La-140 Tritium :f6oo-*- - --- ._____
* *** 427.7 h._...______ ..*.____... _._ ..
Station 87 ' 442.0" "
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- ........... __ ..*.
16 244-70 I Beaufort-Jasper County 265-557 (1 /4) (l1112) Water Treatment Plant (4/4) 76 miles SE
TABLE 3-1 (SHEET 5 of8) 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 Other Control Pathway Total Detectable Locations Annual Mean Stations (g) Locations Sampled Number of Concentration Mean (b), Mean (b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Mean (b), Range Measurement) Performed (Fraction) & Direction Range (Fraction) Range (Fraction) (Fraction) Finished Water Gross Beta 4 3.38 Station 87 3.85 NA 3.00 at Water 48 2.22-6.62 8eaufort-J asper County 2.30-5 .62 1.58-4.36 Treatment (33/36) Water Treatment Plant (11112) (12/13) Plants (pCill) 76 miles SE Gamma Isotopic 48 Be-7 124(e) NDM NDM NA NDM Mn-54 15 NDM NDM NA NDM Fe-59 30 NDM NDM NA NDM Co-58 15 NDM NDM NA NDM Co-60 15 NDM NDM NA NDM Zn-65 30 NDM NDM NA NDM Zr-95 30 NDM NDM NA NDM Nb-95 15 NDM NDM NA NDM 1-131 1 NDM NDM NA NDM Cs-134 15 NDM NDM NA NDM Cs-137 18 NDM NDM NA NDM Ba-140 60 NDM NDM NA NDM La-140 15 NDM NDM NA
..*. _.......... __ ..........._......*...*. __ .. -
NDM Tritium
. 2000---- -- ---
- 454.2
-.. -~- ..- .- -.-....-... ....... -.-.. --.-------.,.....-.----_......_...........
Station 89
~
501.3 NA
..- ......*.
-34:fs-- ----****
16 239-655 Purrysburg Water 303-642 211-476 (9/12) Treatment Plant (3 /4) (2/4) 76 miles SSE
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 Other Control Pathway Total Detectable Locations Annual Mean Stations (g) Locations Sampled Number of Concentration Mean (b), Mean (b), Mean (b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Range Measurement) Performed (Fraction) & Direction Range (Fraction) Range (Fraction) (Fraction) Anadromous Gamma Fish Isotopic (pCi/kg-wet) 0 8e-7 655(e) NA NA NA NA Mn-54 130 NA NA NA NA Fe-59 260 NA NA NA NA Co-58 130 NA NA NA NA Co-60 130 NA NA NA NA Zn-65 260 NA NA NA NA Cs-134 130 NA NA NA NA Cs-137 150 NA NA NA NA Fish Gamma (pCi/kg-wet) Isotopic 7 Be-7 655(e) NOM NOM NA NOM Mn-54 130 NOM NOM NA NOM Fe-59 260 NOM NOM NA NOM Co-58 130 NOM NOM NA NOM Co-60 130 NOM NOM NA NOM Zn-65 260 NOM NOM NA NOM Cs-134 130 NOM NOM NA NOM Cs-137 150 34.0 Station 85 34.0 NA 27.54 28.0-39.6 4.3 miles ESE 28.0-39.6 24.1-30.8 (4/4) (downstream) (4/4) (3/3)
TABLE 3-1 (SHEET 7 of 8) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL
SUMMARY
VogtIe Electric Generating Plant, Docket Nos. 50-424 and 50-425 Burke County, Georgia Medium or Type and Minimum Indicator Location with the Highest Other Control Pathway Total Number Detectable Locations Annual Mean Stations (g) Locations Sampled of Analyses Concentration Mean (b), Mean (b), (MDC) (a) Range Name Distance Mean (b), Mean (b), (Unit of Performed Range Measurement) (Fraction) & Direction Range Range (Fraction) (Fraction) (Fraction) Sediment Gamma (pCi/kg -dry) Isotopic 4 1 - - - - - - - - - 1.-.-- .-- .. -........... .-... .... ---...--.-..... --- ---..-.. ---.-.. -...- ....-.-.-... --.-----... .-... -...--.... --.--....... ----..-.-..-.-..-....-........--.-.. - - ... .--..- ..-.........-.-.........-.---..- .. f--------:- - - - - - - - - j. --- .. --.---.... --....--..--..- - -.-.. . Be-7 655(e) 362.7 Station 83 362.7 NA 296.1 292.5-432.9 0.8 miles ENE 292.5-432.9 230.8-I-----=------,~----f .. -..---..- .... ---.... ___ ._ _____ {~gJ_____ _____ _(Q9~_l!~_!T~~_rp1. __.J~g)---- ---. - -f__c_::_~---f ) §J]_Qg1 _____. . Co-60 70(e) NOM NA NOM NA NDM 1-----=------,--."..---,-------1-.-.. - -...-.-----..-.. -......... -.--..--... --- -..... -.---- -........-... -..-......-... --...... -.-. .........- ..-... -...... --.. -.....--.-.-......-....... --........ .-.-......-....-...-.- ....-..... --.......-... -.....--.--... -.-- .~---------j -...- ...- -..-.-... --.. ---....-.....- ---....- .. - .. . Cs-J34 150 NOM NA NOM NA NOM Cs-13 7 -186----*---- -69.T* .....-.---.---...... Sta tion'Ef3 --- -69.T ---- "'---'--"'I---N -A ---~ 65".-6--- -----* -- ----- 54.0-84.3 0.8 miles ENE 54.0-84.3 54.0-75.9 (2/2) (downstream) (2/2) (2/2)
TABLE 3-1 (SHEET 80f8) 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.
- g. "Other" stations, as identified in the "Station Type" column of Table 2-2, are "Community" and/or "Special" stations.
4.0 DISCUSSION OF RESULTS Included in this section are evaluations of the laboratory results for the various sample types. The Minimum Detectable Difference compares the lowest significant difference between a control station and an indicator station, or the control station and the community station, that can be determined statistically at the 99% Confidence Level (CL). The MOD was determined using the standard Student's t-test. MOD as a tool can quantify plant Vogtle's impact on the sun-ounding environment. A difference in the mean values which was less than the MOD was considered to be statistically indiscernible. The 2011 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 (NOM) were observed or periods for which values were not applicable (e.g., milk indicator, etc.) are listed as NOM and are plotted in the tables as O's. TabJe 4-1 Minimum Detectable Concentrations (MDC) Analysis Water Airborne Fish Milk Grass or Sediment (pCi/J) Particulate (pCi/kg- (pCi/l) Leafy (pCi/kg) or Gases wet) Vegetation (fCi/m3) (pCilkg-wet) Gross Beta 4 10 H-3 2000(a) Mn-S4 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 I 60 Cs-134 15 50 130 15 60 150 Cs-137 18 60 150 18 80 180 8a-140 60 60 La-140 15 15 (a) If no drinking water pathway exists, a value of 3000 pCi/1 may be used. 4-1
(b) If no drinking water pathway exists, a value of 15 pCi/1 may be used . Table 4-2 Reporting Levels (RL) Analysis Water Airborne Fish Milk (pCi/1) Grass or (pCi/I) Particulate (pCilkg-wet) Leafy or Gases Vegetation (fCilm3) (pCi/kg-wet} H-3 20,000 (a) Mn-54 1000 30,000 Fe-59 400 10000 Co-58 1000 30000 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-140 100 400 (a) This is the 40 CFR 141 value for drinking water samples. Ifno drinking water pathway exists, a value of 30,000 may be used. (b) Ifno drinking water pathway exists, a value of20 pCi/l 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. The most significant nuclear event since Chernobyl occurred at Fukushima Daiichi Nuclear Power Plant after the Tohoku earthquake and tsunami on March 11, 2011. Equipment failures and nuclear meltdowns resulted in radioactivity being released into the atmosphere. Southern Nuclear's three sites (Farley, Hatch, and Vogtle) detected 1-131 in REMP samples for several weeks following the disaster. 4-2
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. 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. One data point was excluded from the data set for failing Chauvenet's criterion. This was an abnormally high value for tritium in the finished drinking water samples. Data exclusions are discussed in this section under the appropriate sample type. 4-3
TABLE 4-3 DEVIATIONS FROM RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM COLLECTION AFFECTED DEVIATION CAUSE RESOLUTION PERIOD SAMPLES 2 nd Quarter OSL Dosimeters Non-representative direct Badges were missing at Replaced OSL dosimeters at CR 632225 /CAR Station #36 radiation data collection time beginning of quarter 206805 WSW 13.9 04/ 10/12 Fish No channel catfish were No channel catfish were ODCM requirement of "At least CR 632219/CAR Station 1532 collected observed one sample (largemouth bass) of 206806 any conunercially or recreationally important species in an area not influenced by plant discharges." was met. During the spring Fish Anadromous fish not Miscommunication between Procedure change pending to spawning season Station 1480 collected sampling organization and increase control of modifications to CR 632219/CAR REMP program owner the REMP sampling processes . 206806 Reinstated sampling requirement until ODCM is changed and approved.
4.1 Land Use Census and River Survey In accordance with ODCM 4. 1.2, a land use census was conducted on November 27 , 2012 to determine the locations of the nearest permanent residence, mi Ik 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. The 2012 survey revealed that the WNW and NW sectors no longer have beef cattle within 5 miles making the closest beef cattle location in the WSW sector farther than the previous closest location.
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 1.4 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 4.3 None 4.9 None SSE 4.7 None 4.7 None S 4.4 None 4.3 None SSW 4.7 None 4.6 None SW 3.1 None None None WSW 2.6 None 2.7 None W 3.4 None 4.4 None WNW 1.9 None None None NW 1.5 None None None 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 th an the one in current use. In 2008 , the controlling receptor was moved to a more conservative location at 1.2 miles WSW. This property was acquired by Georgia Power in 2008. The residents were relocated but this property will potentially be used for contract labor in the future. 4-5
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. A new dairy was started at Girard in 2008 and was added to the REMP. Since control stations are approximately 10 miles greater, this dairy is considered an indicator station. Neither current sampling locations nor the controlling receptor were affected by the 2012 land use census resul ts. A survey of the Savannah River downstream of the plant for approximately 100 miles was conducted on November 20, 2012 to identify any new withdrawal of water from the river for drinking, irrigation, or construction purposes. No new usage was identified. These results were corroborated by checking with the Georgia Department of Natural Resources on January 24, 2013; and the South Carolina Department of Health and Environmental Control on December 14, 2012 . Each of these agencies confirmed that no water withdrawal permits for drinking, irrigation, or construction purposes had been issued for this stretch of the Savannah River. The three water treatment plants used as indicator stations for drinking water are located farther downriver. 4-6
4.2 Airborne As specified in Table 2-1 and shown in Figures 2-1 through 2-3, airbome 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 airbome 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 2012 annual average weekly gross beta activity was 25.9 fei /m3 for the indicator stations. It was 0.7 fCi/m3 greater than the control station average of 25.2 fei /m 3 for the year. This difference is not statistically discemible, since it is less than the calculated MOD of 3.42 fei /m3. The 2012 annual avera:fe weekly gross beta activity at the Girard community station was 26.1 fei /m which was 0.9 fei/m 3 greater than the control station average. This difference is not statistically discemible since it is less than the calculated MDD of 4.27 fei /m 3. 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. 4-7
Figure 4.2-1 Average Weekly Gross Beta Air Concentration r:: ~ 15+-~-r-+~--r-+--r-+~--r-+-~-r-+~--r-+--r-+-1--r-+--r-+~~ -g 10
'-
r:: Q) 10 1--1-- ---+---I-i---I--+--+--I-._-r-- --+--+--+-----1----- '--I--f---'----/--r-+-I----/-/-- o U Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year
!--+-Indicator ---- Control - - Community - - MDC !
4-8
Table 4.2-1 Average Weekly Gross Beta Air Concentration Period Indicator Control Community (fCi/m3) (fCilm3) (fCi/m3) 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 2002 19.9 18.9 18.6 2003 19.4 20.5 18.3 2004 21.6 22.8 21.4 2005 20.5 20.4 19.4 2006 25.5 24.6 24.3 2007 27.3 25.1 26.5 2008 24.0 23 .2 23.7 2009 23.0 22.4 22.5 2010 25.8 24.4 25 .5 2011 25 .8 25.1 24.6 2012 25 .9 25 .2 26.1 4-9
During 2012, 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 pre-operation, 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 MOC for airborne Cs-137 is 60 fCi /m 3 . Also, during pre-operation, 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 analyses of the quarterly composites of the air particulate filters. In 2012, Be-7 was not identified in plant gaseous effluents, but it was identified in 2012 REMP air samples and is, therefore, included in the REMP summary table of the airborne pathway samples. The average Be-7 concentration at the indicator stations was 87 .0 fCilm 3 which was 5.0 fCi/ m 3 greater than the average at the control station (82.0 fCi/m\ The difference is not 3 statistically discernible since it is less than the MOD of 28.6 fCi/ m . The average 3 3 at the Girard station was 83 .8 fCi /m which was 1.8 fCi/m greater than the average at the control station. The difference is statistically indiscernible since it is less than the MOD of 51.8 fCi/m 3 . Be-7 has been detected in gaseous effluents in nine of the years of plant operation prior to 2012. However, there was not a statistically discernible difference between the indicator and control station Be-7 concentrations in air samples in 2012 or in any of the years. Airborne 1-131 was not detected in air samples during 2012. During pre-operation, positive results were obtained only during the Chemobyl incident when concentrations as high as 182 fCi/m 3 were observed. The nuclear accident at Fukushima Oaiichi Nuclear Power Plant which occurred after the Tohoku earthquake and tsunami on March 11, 20 II released radioactivity into the environment that was detected in Vogtle air samples. Iodine-131 was detected in air cartridges after Fukushima but no changes in gross beta activity were seen 3 during that same time period. Iodine-131 ranging from 24.7-93.8 fCi /m was seen at Vogtle for several weeks following the Fukushima accident. The MOC and RL 3 for airborne 1-131 are 70 and 900 fCi /m , respectively. 4-10
4.3 Direct Radiation In 2012, direct (external) radiation was measured with Landauer InLight optically stimulated luminescent (OSL) dosimeters which replaced the Panasonic thermoluminescent dosimeters (TLOs). The Panasonic system was retired at the end of2010 due to the inability to keep the aging badge readers operating reliably. Similar to the TLD protocol of the past, two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges. Two direct radiation 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 mi les 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. The mean and range values presented in the "Other" column in Table 3-1 (page 1 of 8) includes the outer ring stations (stations 17 through 32) as well as stations 35 and 43. As provided in Table 3-1 the average quarterly exposure measured at the indicator stations was in 2012 was 14.4 mR with a range of9.6 to 22.7 mR. The average was 0.1 mR greater than the average quarterly exposure measured at the control stations (14.3 mR). This difference is not statistically discernible since it is less than the MOD of 1.5 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 2012 ranged from 9.4 to 20.2 mR with an average of 14.2 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 MOD of 1.5 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 TLOs from Teledyne to Panasonic. It should be noted however that the differences between indicator and control and outer ring values did not change. 4-11
During 2010, OSL badges were co-located on station with the TLD badges. In 2011, only the OSL badges were placed at each station. Following the change to only OSL badges, the differences between indicator, control, and community locations has been consistent with previous years. An increase noted in 20 I 0 reflects issues (especially during 2" Qtr) with the aging Panasonic TLD reader. The close agreement between the station groups supports the position that the plant is not contributing significantly to direct radiation in the environment. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 20 18 16 ~~ ~ V'"' , ~ , 14 \ I \ ~ .
\ ..l ~ ~ . ~
~~I~ ~ ~ ~ r9 y-~ .
C? 12 ~ E -; 10 IJl 8 8 6 4 2 o Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year
-+-Indicator - - - Control - . - Outer Ring I 4-12
Table 4.3-1 Average Quarterly Exposure from Direct Radiation Period Indicator Control Outer Ring (mR) (mR) (mR) Pre-o~ 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 12.9 13.0 12.9 2002 12.8 12.9 12.6 2003 12.2 12.5 12.4 2004 12.4 12.2 12.3 2005 12.5 13.2 12.9 2006 13.1 12.9 13 .0 2007 13.0 12.5 12.7 2008 13 .3 13.0 13.1 2009 13.1 13.6 13.3 2010 16.2 16.7 16.6 2011 13.9 13.9 14.0 2012 14.4 14.3 14.2 4-13
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 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 25 20 /", LI\. v:/~ .J~~
' /k~~'~ I)~ ~
~ 15
'~ v) ~ V V I--I y \~
- E Q) fIl
'- - '~ r--'
-,.. ..-~ >,- - ~- --, " - '
- l--~A
'11'-- .... ,.....
".......... --r -'
o C 10 5 o Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year I-+-- Hunting Cabin (Sta 33) 41- Girard (Sta 35) Rec Center (Sta 43) I 4-14
Table 4.3-2 Average Quarterly Exposure from Direct Radiation at Special Interest Areas Period Station 33 Station 35 Station 43 (mR) (mR) (mR) Pre-op 16.6 IS. 1 IS.3 1987 21.3 18.S lS.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 .S 12.0 1993 12.9 13.3 12.1 1994 12.6 13 .6 12.0 1995 13.3 13.S 12.3 1996 13 .0 13.6 12.1 1997 13 .8 14.4 12.7 1998 13.S 13.7 12.S 1999 NA 14.S 12.7 2000 NA 14.8 13 .1 2001 NA 14.0 12.6 2002 NA 14.0 12.1 2003 NA 14.1 12.2 2004 NA 14.2 1l.7 200S NA lS.2 12.7 2006 NA 14.3 ]2.6 2007 NA 13.6 11.8 2008 NA 14.4 12.6 2009 NA 14.6 13.0 2010 NA 18.0 16.4 20 II NA IS.6 13.6 2012 NA lS.8 14.1 The hunting cabin activities at Station 33 have been discontinued and , consequently, this location is no longer consjdered as an area of special interest. Monitoring at this location was discontinued at the end of 1998. Table 4-3 lists the REMP Program deviations that occurred in 2012 . There was one deviation involving OSL dosimeters. At second quarter collection, badges from Station #36 were missing in the field at the time of collection. The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self imposed limit of 3.S. Previously with TLDs, this limit had been 1.4. However, the OSL readings varied more (between the two elements) than the TLD readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication ISO, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 4-15
1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 20 I O. This limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater than or equal to 3.5 are excluded since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 2012, the OSL results from the following stations were excluded from the data set because their standard deviations were greater than or equal to 3.5 : First Quarter: None Second Quarter: None Third Quarter: None Fourth Quarter: None No badges at any station exhibited a standard deviation greater than or equal to 3.5 in2012. 4-16
4.4 Milk In accordance with Tables 2-1 and 2-2, milk samples are collected bimonthly from two locations, the Girard Dairy (Station 101) which is considered an indicator station because it is approximately 5.5 miles from Vogtle (ideally a milk indicator station is less than 5 miles from the plant), and the Seven Oaks Dairy (Station 102) at 7.5 miles from Vogtle is the control location (ideally control locations are greater than 10 miles from the plant). As discussed in Section 4.1 , no milk animal was found during the 2012 land use census. There were no milk sampling deviations in 2012. Gamma isotopic and I-131 analyses are performed on each milk sample. No Cs-137 was detected in milk samples in 2012. 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 Average Annual Cs-137 Concentration in Milk 20 18 16 1\ S 14 1\ 1\ () Co
~ 12 I
o I \
-
- 10
~ ~
u 8 1\ I \ c: )J-I o 6 () 4 2 o Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year
-+-Indicator - - - Control -- - MOC 4-17
Table 4.4-1 Average Annual Cs-137 Concentration in Milk Year Indicator Control (PCilD (PCilll Pre-op 18.5 18 1987 NOM lOA 1988 NOM 6.9 1989 NOM 7 1990 NOM 17 1991 NOM 14.2 1992 NOM NOM 1993 NOM NOM 1994 NOM NOM 1995 NOM NOM 1996 NOM NOM 1997 NOM NDM 1998 NOM NOM 1999 NOM NOM 2000 NOM NOM 2001 NOM NOM 2002 NOM NOM 2003 NOM NOM 2004 NOM NOM 2005 NOM NOM 2006 NOM NOM 2007 NOM NOM 2008 NOM NOM 2009 NOM NOM 2010 NOM NOM 2011 NOM NOM 2012 NOM NOM Following the Fukushima accident which began on March 11,2011,1-131 was detected in milk samples collected from both dairy locations. Positive l-131, ranging from 1.63 to 12.5 pCi/I, was seen in the late March samples and in the samples from both collection dates in April. No other samples from 2011 contained 1-131. Since plant 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. Ouring preoperation, positive 1-131 results were found only during the Chemobyl incident with concentrations ranging from 0.53 to 5.07 pei/I. The MOC and RL for 1-131 in milk are 1 and 3 pCill, respectively. 4-18
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 WSW from the plant (Station 37). Gamma isotopic analyses are performed on the samples. The man-made radionuclide Cs-137 has been found in vegetation samples in the past, however, there was no Cs-137 detected in any of the 2012 vegetation samples. Cesium-137 is often detected in environmental samples as a result of atmospheric weapons testing and the Chernobyl incident. The naturally occurring radionuclide Be-7 is typically detected in indicator and control station vegetation samples. Be-7 was not detected in gaseous effluents in 2012, but it was detected in REMP vegetation samples and is, therefore, included in the REMP summary table of the airborne pathway samples. The average at the indicator stations was 2724.6 pCi/kg-wet which is 1025.5 pCi/kg-wet higher than the average at the control station (1699.1 pCi/kg-wet). The difference between the averages at the indicator and control stations is not statistically discernible since it is less than the MOO of 2679 pCi/kg-wet. Be-7 has been detected in gaseous effluents in nine of the previous 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. There is no Required MOC or Reporting Level for Be-7. The historical trending of the average concentration of Cs-13 7 at the indicator and control stations is provided in Figure 4.5-1 and listed in Table 4.5-l. No trend is recognized in this data. The MOC and RL for Cs-13 7 in vegetation samples are 80 and 2000 pCi/kg-wet, respectively. Cs-l37 is the only man-made 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. 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. 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 MOC or RL for Co-60 in vegetation. In 2006, one sample at the indicator station was positive for Cs-l37 at a higher concentration, 49l.8 pCi/kg-wet, than typically seen over the years in Vogtle vegetation samples. A duplicate sample (which is taken periodically) happened to be taken at the same collection time and also revealed a similar activity. The higher concentration more than likely resulted from plowing and seeding activities (to maintain the vegetation plot) which took place a couple of weeks prior to the sample collection. 4-19
In 2011, following the nuclear accident at Fukushima Daiichi Nuclear Power Plant, 1-131 was detected in REMP vegetation samples. Iodine-131 was detected at Vogtle in all three forage samples collected on 03 /29111 (after the Fukushima event), but not in any other monthly forage samples collected in 2011. The range of 1-131 values was 58.3 to 81.8 pCi/kg-wet. The MDC and RL for 1-131 in vegetation are 60 and 100 pCilkg-wet, respectively. During 2012, 1-131 was not detected in any Vogtle REMP vegetation samples. 4-20
Table 4.5-1 Average Annual Cs-137 Concentration in Vegetation Year Indicator Control lQCi/k~-we!l ~Ci/~-we!l Pre-op 54,6 43 ,7 1987 24.4 61.5 1988 38 ,7 NOM 1989 9,7 NOM 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 NOM NOM 1997 NOM 32,6 1998 NOM 50,1 1999 37,2 NOM 2000 36,6 NOM 2001 NOM NDM 2002 NOM 98,3 2003 24,S NDM 2004 36,8 19,7 2005 49,S NOM 2006 23,9 491.8 2007 20,2 NDM 2008 24,6 62,1 2009 34,6 NOM 2010 NDM NDM 2011 69,6 NDM 2012 NDM NDM 4-22
4.6 River Water Surface water from the Savannah River is obtained at three locations usmg 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 2012 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 1488 pCifl which was 1076 pCill greater than the average at the control station (412 pCill). MOD was not calculated because only one positive result occurred at the Control Station. For all other results tritium was at non-detectable levels. The MOC for tritium in river water used to supply drinking water is 2000 pCifl and the RL is 20,000 pCifl. At the special river water sampling station, the results ranged from 258 pCifl to 636 pCifl with an average of 446 pCifl. MOD was not calculated because only one positive result was occurred at the Control Station. For all other results tritium was at non-detectable levels. The decrease in tritium concentration between the indicator station and the special station is due to the additional dispersion over the 0.9 miles that separates the two stations. In the first two years of operation, the tritium concentration at the special station was somewhat greater than that at the indicator station. In recent years, the level at the special station has generally become less than the level at the indicator station. 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 MOD between the indicator and control stations; and the total curies of tritium released from the plant in liquid effluents. The annual downriver survey of the Savannah River, as discussed in Section 4.1, indicated that river water is not being used for purposes of drinking or liTigation for at least 100 miles downriver. 4-23
Figure 4.6-1 Average Annual H-3 Concentration in River Water 3000 2500 J -- V1\ \ U
-3; c
0 2000
- 1500
\
\ / i\ 1/ \ I \ / \ I1\
-....
('C c Q) 0
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Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year
!-+-Indicator -II- Control - - Special - MDC I 4-24
Table 4.6-1 Average Annual H-3 Concentration in River Water Year Special Indicator Control Difference MDD Annual Site (pO/I) (pO/I) (pCi/1) Between (pO/I) Tritium Indicator and Released Control (Ci) (pCi/I) 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 2002 1280 2628 437 2190 1211 1566 2003 800 1376 399 977 1706 1932 2004 743 1269 351 918 1061 1212 2005 713 800 458 342 1333 1860 2006 852 2307 384 1882 2688 2005 2007 489 879 344 535 1189 757 2008 1105 1874 832 1042 4838 1364 2009 614 1203 221 982 3551 1224 2010 607 814 235 579 2094 903 2011 880 2068 409 1659 2522 1361 2012 446 1488 412 1076 NA 1810 4-25
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 three indicator locations (Station 87 - the Beaufort-Jasper County Water Treatment Plant near Beaufort, South Carolina, 112 river miles downriver; Station 88 - the Cherokee Hill Water Treatment Plant near Port Wentworth, Georgia, 122 river miles downriver; and Station 89 - the Purrysburg Water Treatment Plant near Purrysburg, South Carolina, located about 112 miles downriver. The Purrysburg Station was added to the REMP in January 2006.) Stations 87 and 89 are located on the same canal with the Purrysburg location at the beginning of the canal (nearer the Savannah River) and the Beaufort-Jasper location near the end of the canal. 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 2012, the indicator station average gross beta concentration in the raw drinking water was 3.38 pCi/1 which was 0.7 pCi/1 greater than the average gross beta concentration at the control station (2 .65 pCi/I) . This difference is not statistically discernible since it is less than the calculated MDD of 1.05 pCi /1. Through the years, there has been close agreement between the gross beta values at the indicator stations and the control station which supports that there is no significant gross beta contribution from the plant effluents. The required MDC for gross beta in water is 4 .0 pCi/1. There is no RL for gross beta in water. For 2012, the indicator station average gross beta concentration in the finished drinking water was 3.13 pCill which was 0.13 pCi/1 more than the average gross beta concentration at the control station (3.00 pCi/I). This difference is not statistically discernible since it is less than the MDD of 0.87 pCi/1. The gross beta concentrations at the indicator stations ranged from 0.79 to 5.50 pCi/1 while the concentrations at the control station ranged from 1.58 to 4.36 pCi/1. The required MDe for gross beta in water is 4.0 pCi/ 1. There is no RL for gross beta in water. 4-26
Figure 4.7-1 Average Monthly Gross Beta Concentration in Raw Drinking Water 8 7
- 6 Ga. I
-s 1:\ ~ c: o 4 I \\ /~ \ IV \
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U ') o Po 87 88 89 90 ~ 92 93 94 9S 96 97 98 99 00 01 02 03 04 OS 06 07 08 09 10 11 12 Year
-+--Indicator ---- Control - - - MOC I 4-27
Table 4.7-1 Average Monthly Gross Beta Concentration in Raw Drinking Water Period Indicator Control
~CU!l (QCi/!l 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 2002 3.09 2.61 2003 3.73 2.59 2004 4.06 2.39 2005 3.75 2.48 2006 3.85 2.93 2007 4.00 3.13 2008 3.46 2.37 2009 3.28 2.26 2010 2.95 1. 71 2011 2.31 2.26 2012 3.38 2.65 4-28
Figure 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water 4.5 4 - ----- 1-- e--' 1--- - -- -- -- -
- 3.5 U .L~
..---'
'\ )~ I\--.. J~
~ c: o 3 2.5 \ )~I\ I / V. ~ VI'\ \ \~ ~V rx I
\ ki J::=I ~ p 'N k"-' / \ V l~ II \~ ~
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11-c: o U 0.5 o Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year I -+-Indicator ----- Control - - MOC I 4-29
Table 4.7-2 Average Monthly Gross Beta Concentration in Finished Drinking Water Period Indicator Control (pCi/I) (pCi/I) 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 2002 2.80 2.61 2003 2.51 2.34 2004 2.36 1.92 2005 2.61 2.00 2006 3.23 3.25 2007 3 .19 3.36 2008 2.86 2.07 2009 2.53 2.13 2010 2.89 2.23 2011 2.36 3.13 2012 3.13 3.00 4-30
As provided in Table 3-1, there were no POSitive results during 2012 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/1 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 MOC assigned for Be-7 in water is 124 pCi /1. Also during preoperation, Cs-134 and Cs-13 7 were detected in about 7% of the samples at concentrations on the order of their MOCs which are 15 and 18 pCi/l, respectively. I-131 was detected in finished drinking water in 1997 at levels near the MDC. This was the first occurrence for detecting I-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 pCill at Port Wentworth. The MDC and RL for I-131 in drinking water are 1 and 2 pCi/l, 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 MODs (when applicable) 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 2012 raw drinking water indicator stations average tritium concentration was 428 pCi/1 which was 218 pCi/1greater than the single positive concentration found at the control station (210 pCi/l). MDO was not calculated between the indicator and control stations due to only one positive result at the control station. For all other results tritium was at non-detectable levels. The MOC and RL for tritium in drinking water are 2000 pCi/1and 20,000 pCi/l, respectively. The finished drinking water average tritium concentration at the indicator stations during 2012 was 454 pCi/1 which was 111 pCi /1 greater than the average concentration found at the control station (344 pCi/l). This difference is not statistically discernible since it is less than the calculated MOD of 357 pCi/i. One data point was excluded from the finished drinking water data set due to failing Chauvenet's Criterion. No reason for the abnormally high result could be determined. The MOC and RL for tritium in drinking water are 2000 pCi/1 and 20,000 pCi /I, respectively. 4-3 J
Figure 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water 3000 2500 .......... r---. / <.> 2000 \ -o a. t: -g ~ 1500
...
\/
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t: (l) 1000 :\ "--., ~ ..... ...... v
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Po 87 88 89 90 91 '" 92 93 94 95 96 97 98 99 00 01 Year 02 03 04 05 06 07 08 09 10 11 12
!--+--Indicator - - - Control - MDC I 4-3 2
Table 4.7-3 Average Annual H-3 Concentration in Raw Drinking Water Period Indicator Control Difference MDD (pC ill) (pCill) Between Ind. & (pCi/I) Control (pCiJl) 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 l373 179 1194 353 1993 955 NDM 955 NA 1994 871 NDM 871 NA 1995 917 201 716 NA 1996 1014 207 807 151 1997 956 230 726 61 1998 791 160 631 NA 1999 908 NDM 908 NA 2000 1020 373 647 704 2001 889 525 364 NA 2002 938 304 634 284 2003 563 203 360 NA 2004 585 220 365 204 2005 463 393 70 301 2006 690 451 239 394 2007 462 357 105 NA 2008 726 386 340 269 2009 602 587 15 NA 2010 343 244 99 205 2011 464 211 253 NA 2012 428 210 218 NA 4-33
Figure 4.7-4 Average Annual H-3 Concentration in Finished Drinking Water 3500 3000 22500 1\ / 1\\ U -Co t: o 2000 1\
E 1: 1500
\V
~
Q) (J t: 8 1000 500
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Po 8788899091929394 95 96 9L98 99 00 0102 03 04 05 06 07 08 09 10 1112 Year
---+-Indicator ---- Control - - MOC I 4-34
Table 4.7-4 Average Annual H-3 Concentration in Finished Drinking Water Period Indicator Control Difference MDD (pO/I) (pO/I) Between Ind. & (pO/I) Control JpCi/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 NOM 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 2002 1060 340 720 416 2003 473 196 277 NA 2004 531 255 276 314 2005 546 223 323 NA 2006 688 710 22 NA 2007 494 229 265 NA 2008 661 391 270 468 2009 579 667 88 NA 2010 374 262 112 NA 2011 410 226 184 NA 2012 454 344 110 357 4-35
4.8 Fish Table 2-1 requires 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 semi-annual 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. Anadromous fish were not sampled in 2012. The sample was not taken due to a miscommunication between the sampling organization and the REMP program owner. Direction was given to the sampling organization to cease the sampling of the anadromous fish species. A change to the ODCM had been planned but did not get implemented when the REMP coordinator position was changed in 2012. Procedure changes are pending to increase the control of modifications to REMP sampling processes to prevent recurrence (CR 632219). The bases for discontinuing this sample are being reviewed by the current REMP coordinator and this sample may be discontinued in the future following an approved change to the ODCM. In all but three previous years of operation, no radionuclides were detected in anadromous fish samples. In 2005, Cs-137 was detected in the anadromous fish sample at a low level of28.8 pCi/kg-wet. 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 semi-annual catches at the indicator and control stations during 2012 are shown below. As indicated in Table 4-3, Channel catfish were not observed during the spring sample collection and, therefore, no sample of this species was collected. Date Indicator Control April 10 Largemouth Bass Largemouth Bass April 10 Channel Catfish November 2 Largemouth Bass Largemouth Bass November 2 Channel Catfish Channel Catfish As indicated in Table 3-1, Cs-137 was found in the semiannual collections of a commercially or recreationally important species of fish and in fish at the control station. It has been found in all but 5 samples collected during operation and in all but 5 of the 32 samples collected during preoperation. As provided in Table 3-1, 34.0 pCi/kg-wet was the average Cs-137 detected in the 4 samples from the indicator station, and 27.5 pCi/kg-wet was the average Cs-137 detected at the control station. The difference of is not statistically significant since it is less than the MDD of 14.1 pCi/kg-wet. No discernible difference between the indicator and control stations has OCCUlTed for any year of operation or during pre-operation. 4-36
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. The indicator station fish sample concentration of Cs-13 7 in 1999 was greatly influenced by a largemouth bass collected in October with a concentration of 2500 pCi/kg-wet. Other than the fact that largemouth bass are predators that concentrate Cs-137, no specific cause for the elevated concentration in this sample is known. No trend is recognized in this data. The MDC and RL for Cs-137 in fish are 150 and 2000 pCi/kg-wet, respectively. Figure 4.8-1 Average Annual Cs-137 Concentration in Fish 900 800 ~700
~I
~600 U 1\ S500 s::: o
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I'::-J /. ~ o \.- K ~ Po 87 88 89 90 91 92 93 94 95 96 9Ye~~ 99 00 01 02 03 04 05 06 07 08 09 10 11 12
--+- Indicator - - - Control - - MOC 1 4-37
Table 4.8-1 Average Annual Cs-137 Concentration in Fish Year Indicator Control (pCi/kg-wet) (pCi/k2-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 2002 59 133 2003 62 21 2004 56.4 26.0 2005 39.3 40.2 2006 257 35.7 2007 58.7 37.7 2008 39.4 47.0 2009 NOM 30.4 2010 42.8 74.4 2011 42.6 30.5 2012 27.5 34.0 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 pCilkg-wet, respectively. The MDC assigned to 1-131 in fish is 53 pCi/kg-wet. In the November 2008 collection, the Largemouth Bass sample from the control location showed 90 pei/kg-wet of 1-131. The specific source of the 1-131 is unknown but is likely due to medical waste. During preoperation, Cs-134 was found in two of the 17 samples collected at the control station at concentrations of 23 and 190 pei/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. 4-38
4.9 Sediment Sediment was collected along the shoreline of the Savannah River on April 3 and October 2, 2012 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 radionuclides of interest identified in 2012 samples were Be-7 and Cs-137 . Be-7 , which is abundant in nature, was not identified in plant liquid effluents during 2012. It continues to be trended in river sediment in the REMP report even when not identified in plant effluents . In 2012, the average Be-7 concentration at the indicator station was 363 pCi/kg-dry and at the control station the average concentration was 296 pCi/kg-dry. This difference (67 pCi /kg-dry is less than the MDD (95l.5 pCi/kg-dry) and is therefore not statistically discernible. Due to the low number of samples, the variability of Be-7 activity found in them, and the high standard deviation the MOD value is very high. Because Be-7 is often not identified in plant effluents and because there is no significant difference between the indicator and control station, the Be-7 found at the indicator station is not attributed to plant releases . For Cs-137, the average concentration at the indicator station during 2012 was 69.1 pCi/kg-dry which was 4.1 pCi/kg-dry greater than that at the control station (65.0 pCi/kg-dry). The difference between the average value at the indicator station and the average value at the control station is not statistically discernible since it is Jess than the calculated MOD of 185.4 pCi/kg-dry. However, the concentration of Cs-137 found at the indicator station could be attributed to plant effluents or to other facilities that release radioactive effluents in the vicinity of the plant. The Cs-137 level at the indicator station has averaged nearly 100 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 . During 2012, Co-60 was not detected in any of the four sediment samples. Cobalt-60 has been detected in sediment collected at the indicator station every year of plant operation but four. The concentrations of Co-60 often found at the indicator station could be attributed to plant releases or, potentially, to other facilities that release radioactive effluents in the vicinity of the plant. 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-13 7) 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 4-39
respective MOC values. The presence of these preoperational nuclides could be attributed to atmospheric weapons testing and the Chemobyl incident. Mn-54, 1-131, and Cs-134 have been found sporadically during the years of operation. A summary of the positive results for these nuclides along with their applicable MDCs is provided in Table 4.9-5. Figure 4.9-1 Average Annual 8e-7 Concentration in Sediment 3500 I 3000 1\ F "0 en 2500 I\ ~ U S 2000 I \ I V \ / 1\ I~\
~
t: 0
- 1500 /\ !. \
-co
~
t:
/1\ II \ 1'"'If! ;\ \ V l\ if
'/ \\/~ ..---.... \N ~ lr(/) /
~
J
'"
(1) u 1000 t: 0 ()
/ ./1\. \ IL' / !.
V
"\ !j '\V 1\"- \
500 r- "- ..... / \
~ ' t J--I 0
Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year I -+-Indicator ---- Control - - MOC ! 4-40
Table 4.9-1 Average Annual Be-7 Concentration in Sediment MDC=655 pCilkg-dry Year Indicator Control JpCi/kg-dry) lpCi/kg-dry) Pre-op 580 500 1987 987 543 1988 970 810 1989 1300 415 1990 465 545 1991 826 427 1992 2038 380 1993 711 902 1994 1203 964 1995 1865 1575 1996 1925 831 1997 1130 1028 1998 1396 1016 1999 662 769 2000 1526 3324 2001 1697 2614 2002 742 1254 2003 1150 903 2004 1309 905 2005 1931 1086 2006 1254 704 2007 1034 1274 2008 394 805 2009 2011 1131 2010 1217 533 2011 885 380 2012 363 296 4-41
Figure 4.9-2 Average Annual Co-58 Concentration in Sediment 300 ~ 250 "0I Cl ~ 200 U -gQ. 150 1\
~
....c: 100 j ~
;\
I \ I1\ Q) (.) c: o U 50 o
- 1-- -
1\17- nV L
-- . - .. - -- -- '--- .
Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year
-+-Indicator - - - Control --. MOC I 4-42
Table 4.9-2 Average Annual Co-58 Concentration in Sediment MDC=43pCi/k~-dry Year Indicator Control (pCilkg-dry) (pCi/kg-dry) Pre-op NOM NOM 1987 NOM NOM 1988 190 NOM 1989 135 NOM 1990 140 NDM 1991 NDM NDM 1992 124 NDM 1993 NOM NOM 1994 18.4 NOM 1995 42.4 NOM 1996 274 NDM 1997 NOM NOM 1998 NDM NOM 1999 NDM NOM 2000 NDM NOM 2001 NDM NOM 2002 NOM NOM 2003 NOM NOM 2004 NOM NOM 2005 NDM NOM 2006 NOM NOM 2007 NDM NDM 2008 NDM NOM 2009 NOM NOM 2010 NOM NDM 2011 NOM NDM 2012 NOM NOM 4-43
Figure 4.9-3 Average Annual Co-50 Concentration in Sediment 400 350 ~ / "0 300 ~ I Cl V U 250 - c.. c:: 200 0 I n
10
.... 150
+' c:: I I \ u Q) c:: 100 1/ \1/
\ / /1\ / 1\
U 0
/ \ ./ / \ II /
\
50 ~ V ~ \ \ 0 V, \V ~/ \ Po 87 88 89 90 9192 93 94 95 96 9L98 99 00 0102 03 04 05 06 0708 09 10 1112 Year
-+--Indicator ---- Control - - MOC I 4-44
Table 4.9-3 Average Annual Co-60 Concentration in Sediment MDC=70 pCilkg-dry Year Indicator Control (pCilkg-dry) (pCilkg-dry) Pre-op NOM NOM 1987 NOM NOM 1988 62 NOM 1989 46 NOM 1990 46 NOM 1991 113 NOM 1992 59.5 NOM 1993 65.9 NOM 1994 85.2 NOM 1995 267 NOM 1996 344 NOM 1997 86 NOM 1998 263 NOM 1999 49.5 NOM 2000 131.3 NOM 2001 NOM NOM 2002 49.7 NOM 2003 146 NOM 2004 77 NOM 2005 146 NOM 2006 40 NOM 2007 NOM NOM 2008 61.9 NOM 2009 NOM NOM 2010 NOM NOM 2011 NOM NOM 2012 NDM NOM 4-45
Figure 4.9-4 Average Annual Cs-137 Concentration in Sediment 600 500 I ~
~
'UI 400 II
/\
-g \ U c.. 300 /\ /
C1l
'- ~
u 200
\
" /
J\
"\
/
~
/ \/
V\ V j 1\ j \ U o c:
~
~V .-' I"-. """- / '"'-- ~
1\ r----.. \ r--. ~ /, V 1\
...... , ..J~ f/ "'-.
100
\V ~
o Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 Year I -+--Indicator - - - Control - - - MDe I 4-46
Table 4.9-4 Average Annual Cs-137 Concentration in Sediment MDC=180 pCilkg Year Indicator Control (pCilkg) (pCilkg) Pre-op 320 150 1987 209 III 1988 175 175 1989 230 125 1990 155 140 1991 246 100 1992 259 III 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 2002 189 60 2003 171 90 2004 149 100 2005 263 89 2006 142 68 2007 125 83 2008 66.2 60.9 2009 127.7 103 .2 2010 164.6 64 .1 2011 85.1 63 .1 2012 69.1 65.0 Table 4.9-5 Additional Sediment Nuclide Concentrations Nuclide YEAR Indicator Control MDC (pCilkg-dry) (pCilkg-dry) (I!Ci/kg-dry) Mn-54 1988 22 NDM 1989 18 NDM 42 1994 32 NDM 1-131 1992 194 20 53 1994 51 41 Cs-134 2011 18 NDM 150 4-47
4.10 Groundwater As nuclear plants began to undergo decommissioning in the late 1990s to early 2000s, instances of subsurface and/or groundwater contamination were identified. In addition, several operating facilities also identified groundwater contamination resulting from spills and leaks or equipment failure. In one instance, low levels of licensed material were detected in a private well located on property adjacent to a nuclear power plant. In 2006, NEI (Nuclear Energy Institute) fonned a task force to address monitoring onsite groundwater for radionuclides at nuclear facilities. A Groundwater Protection Initiative was developed which was adopted by an U.S. commercial operating nuclear plants. The NRC also fonned a task force to study the groundwater issues and released Information Notice 2006-13, Ground-water Contamination due to Undetected Leakage of Radioactive Water, which summarized its review of radioactive contamination of ground water at multiple facilities as a result of undetected leakage from structures, systems, and components that contain or transport radioactive fluids. Licensees were instructed to review the infonnation for applicability and to consider appropriate actions to avoid similar problems. The NEI task force felt it was prudent for the industry to update site hydrology infonnation and to develop radiological groundwater monitoring plans at each site. These groundwater protection plans would ensure that underground leaks and spills would be addressed promptly. Additionally, the task force recommended developing a communications protocol to report radioactive leaks or spills that entered groundwater (or might eventually enter groundwater) to the NRC and State / Local government officials as needed . NEI-07-07, Industry Groundwater Protection Final Guidance Document, was developed by the task force to document the guidelines recommended for the industry. To ensure compliance with NEI-07-07, Southern Nuclear developed the Nuclear Management Procedure, Radiological Groundwater Protection Program. The procedure contains detailed site-specific monitoring plans, program technical bases, and communications protocol (to ensure that radioactive leaks and spills are addressed and communicated appropriately). In an effort to prevent future leaks of radioactive material to groundwater, SNC plants have established robust buried piping and tanks inspection programs. 4-48
In 2006, Vogtle sampled onsite drinking water deep wells and onsite makeup water deep wells for tritium and gamma isotopic activity. These wells did not contain detectable amounts of radioactivity. In 2007, Vogtle implemented a more extensive radiological groundwater monitoring program. A qualified hydrologist made recommendations for drilling additional onsite monitoring wells and updated the site hydrology information. Eight new wells and 17 existing wells comprise the current VEGP groundwater monitoring program (see Table 2-3). These locations were sampled twice in the latter portion of 2007. Several wells were positive for tritium but no gamma emitters were detected . The highest activity sample showed approximately 900 pCi/1 of tritium. This level of tritium is typical background for the area around Plant Vogtle based on historical information from Georgia Department of Natural Resources and Savannah River Site. Drinking water wells, sewage treatment plant effluent, and several surface water locations supplement the monitoring program and were also sampled in 2007. None of these locations showed activity above typical environmental levels in this area. This is also true of the 2008 supplemental sampling. The tritium levels in the water table since the radiological groundwater sampling program started in mid-2007 through 2012 are graphed in Figures 4.10-1, 4.10-2, and 4.10-3. The February 2008 sampling event appears to be an outlier, however, more data is needed to determine seasonal changes and typical fluctuations in tritium concentration due to rain washout and recharge of the aquifer. None of the tertiary aquifer wells have shown tritium concentrations above background. In 2008, three of the monitoring wells (1013, 1003, and 1004) used for groundwater monitoring (but not newly drilled for the program) were retired due to preliminary construction activities of two potential new operating reactors. These wells were not critical to the radiological groundwater monitoring program as they were upgradient and used primarily to obtain background data for site characterization. In 2009, upgradient well 805-A had "silted in" and is now only being used for groundwater level data. In 20 I 0, tertiary aquifer wells 27 and 29 were no longer sampled due to structural issues with the wells that made sampling extremely labor intensive. It was determined that enough background data had been gathered from these wells. In 2012, tritium concentrations observed in the Vogtle groundwater monitoring wells fluctuated but did not exceed the established Administrative Control Limits (ACLs). The ACLs were derived based on previous years ' tritium results and total measurement uncertainty and are site specific by plant and aquifer. There are no reporting requirements associated with exceeding an ACL but additional actions would be taken to verify no new sources of tritium were contributing to the increase. For the deeper aquifer sampled at Vogtle, the ACL is 1600 pCi/L and for the surficial aquifer, the ACL is 2100 pCi /L. 4-49
VEGP GW Wells in PA FIGURE 4.10-1 1400 -, 1200 11----------------------~----___ 1000 1 R! "
- ! 800
.t:>- U o. In M 0
- I:
600 400 -tl - - - -- - -- - - - - 200 +1- - - -- -- -- - -- -- o ~!-----------~--~-------------r----------------~------------------.------------------4 Oct-06 Feb-08 Jul-09 Nov-10 Apr-12 Aug-13 C-+-LT-1B --- LT-7A LT-12 - -LT-13]
Existing Water Table Wells (Zeroes are below MDA) FIGURE 4.10-2 1400 ,----------------------------- 1200 I . . , \ 1000 I 1+1-----------------+--+-----------
- !
U """ U, f-' M c..
- I:
600 400 - t"r:--t 200 I H \ I oI ~ ~_,------,_------ r Apr-O? Nov-O? Jun-08 Dec-08 Ju\-09 Jan-10 Aug -10 Feb-11 Sep-11 Apr-12 Oct-12 May-13
-+-802A -+-- 803A 805A 8068 ----*-808 ..... OW-1013 - - OW-1015 - OW-1004
New Water Table Wells (Zeroes are below MDA) FIGURE 4.10-3 1800 1600 1400 - 1200 .,., -ua.
...J 1000 (J'1 N M
- I:
800 600 400 200 o .' * .. e I 9 Nov-O? Jun-08 Dec-08 Jul-09 Jan-10 Aug-10 Feb-11 Sep-11 Apr-12 Oct-12 May-13 [ -+-R-1 _ R-2 R-3 R-4 --*-R*5 -+-R-6 --+-R-? --R-8
5.0 INTERLABORATORY COMPARISON PROGRAM In accordance with ODCM 4.1.3, the EL participates in an ICP that 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 the rcp is to be conducted with the Environmental Protection Agency (EPA) Environmental Radioactivity Laboratory rntercomparison Studies (Cross-check) Program or an equivalent program, and the rcp should include all of the determinations (sample medium/radionuclide combinations) that are offered by the EPA and included in the REMP . The rcp is conducted by Analytics, Inc. of Atlanta, Georgia. Analytics has a documented Quality Assurance (QA) program and the capability to prepare Quality Control (QC) materials traceable to the National Institute of Standards and Technology. The rcp 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 and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses of water samples. 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 as reported by Analytics. The uncertainty of the reported average is the propagated error of the values in the reported average by the EL. The precision of each result is measured by the coefficient of variation, which is defined as the standard deviation of the reported result 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: 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 (pCi/I).
5-1
As required by ODCM 4.1.3 .3 and 7.l.2.3, a summary of the results of the EL's participation in the rcp is provided in Table 5-1 for: the gross beta and gamma isotopic analyses of an air filter; gamma isotopic analyses of milk samples; and gross beta, tritium and gamma isotopic analyses 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. The Environmental Radiochemistry laboratory participates in a performance evaluation (PE) sample program provided by Analytics Inc. The PE samples are received and analyzed routinely with environmental and effluent samples. The laboratory analyzed 9 samples for 35 parameters in 2012. The 2012 analyses included tritium, gross beta and gamma emitting radio-nuclides in different matrices. The attached results indicate 3 analyses (Ce-141, Cr-51, and Fe-59) were outside the acceptance limits for accuracy. These isotopes were in the Gamma in Air Filter matrix. After the results were received, the sample was recounted but two of the isotopes had decayed off. The remaining isotopes were within acceptable limits for accuracy. A Gamma in Air Filter PE sample will be analyzed in 2Q 2013 to complete an investigation. 5-2
TABLE 5-1 (SHEET 1 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS 1-131 ANALYSIS OF AN AIR CARTRIDGE (pCilcartridge) Analysis or I Date I Reported I Value Known , Standard I Uncertainty Percent coef, Normalized Radionuclide Prepared Average Deviation EL Analytics (3S) of Variation Deviation 1-131 1 07/ 14112 1 100.00 I 97.201 6.451 1.62 7.221 OA4 GAMMA ISOTOPIC ANALYSIS OF AN AIR FILTER (pCi/filter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation I Ce-141 09/13112 113.00 153 .00 6.66 2.56 7.72 0.02 I Co-58 09113112 77.50 94.10 3.85 1.57 7.41 -2.90 Co-60 09113112 117.00 142.00 4.14 2.38 5.69 -3.82 Cr-51 09113112 184.00 232.00 18.3 3.88 14.27 -1.84 I , VI Cs-134 09113 / 12 83.40 101.00 2.73 1.69 5.39 -3.92 I w Cs-137 09 / 13112 130.00 163.00 4.83 2.73 5.96 -4.22 Fe-59 09113/12 111.00 142.00 7.18 2.38 8.76 -3.14 Mn-54 09113 / 12 150.00 183.00 12.5 3.06 9.49 -2.29 Zn-65 09/ 13/12 154.00 180.00 13A 3.01 10.16 -2.29
- - - - -- -- - ~ - - - - -- - --- - -- --- -- - - - - - - - - --- . ------~
Analysis or I Date , Reported , Known , Standard ,uncertainty ,percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 1 09113112 1 93.00 I 84.10 J 1.2 I lAol 4.21 2.27
TABLE 5-1 (SHEET 2 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF A MILK SAMPLE (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized RadionucIide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Ce-141 7114112 76.60 82.20 2.85 1.37 8.62 -0.84 Co-58 7114112 90.20 99.70 2.85 1.66 7.33 -0.32 Co-60 7114112 350.00 355.00 10.7 5.93 4.23 -0.35 Cr-51 7114112 433.00 402.00 21.1 6.71 8.51 0.85 Cs-134 7114112 180.00 174.00 6.26 2.91 4.68 0.69 Cs-137 7114112 216.00 212.00 9.26 3.54 5.63 0.34 Fe-59 7114112 126.00 128.00 6.53 2.13 8.21 -0.17 1-131 7/ 14/ 12 102.00 99.70 7 1.66 8.78 0.30 Mn-54 7114/ 12 134.00 132.00 3.79 2.21 5.69 0.24 Zn-65 7114112 208.00 199.00 8.17 3.33 7.09 0.59 VI .h GROSS BETA ANALYSIS OF WATER SAMPLE (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized RadionucIide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Gross Beta 03115112 263.00 297 .00 18.93 4.96 1.10 -11.99 07114112 166.00 148.00 10.28 2.47 10.85 0.98 GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percen t Coef Normalized RadionucIide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Ce-141 03115112 190.00 184.00 5.65 3.07 5.25 0.64 Co-58 03115112 92.50 93.40 5.19 1.56 8.20 -0.11 Co-60 03 / 15112 208.00 197.00 5.68 3.29 4.59 1.16
TABLE 5-1 (SHEET 3 of 3) INTERLABORATORY COMPARISON PROGRAM RESULTS GAMMA ISOTOPIC ANALYSIS OF WATER SAMPLES CONT. (pCi/liter) Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation Cr-51 03 / 15112 362.00 309.00 46.7 5.16 15 .63 0.94 Cs-134 03115112 108.00 106.00 1.75 1.77 4.41 0.49 Cs-137 03 /15112 124.00 113.00 3.67 1.88 6.09 1.50 Fe-59 03 / 15 / 12 119.00 119.00 6.14 1.99 8.23 0.02 1-131 03 / 15112 104.00 93.80 4.15 l.57 6.68 1.44 Mn-54 03115112 149.00 138.00 2.24 2.31 5.13 1.38 Zn-65 03 / 15112 245.00 235.00 5.41 3.93 5.98 0.67 TRITIUM ANALYSIS OF WATER SAMPLES (pCi/liter) , V> V> Analysis or Date Reported Known Standard Uncertainty Percent Coef Normalized Radionuclide Prepared Average Value Deviation EL Analytics (3S) of Variation Deviation H-3 03/15112 4160 4470 102.54 74.70 4.43 -l.70 07/14112 4580 4970 92.36 83.00 4.34 -l.98
6.0 CONCLUSION
S This report confirms the licensee's conformance with the requirements of Chapter 4 of the ODCM. It provides a summary and discussion of the results of the laboratory analyses for each type of sample. In 2012 tritium was detected in finished drinking water samples. Using the ingestion dose factors and consumption rate factors in Reg. Guide 1.109 it was calculated that the highest potential dose to a maximum exposed member of the public (an adult), due to regular consumption of drinking water containing tritium at the low level seen in the 2012 samples (using indicator station average concentration of 454 pCi/l), would be approximately 3.48E-2 mrem in a year. This dose is about 1.2% of the regulatory limit of 3 mrem per year due to liquid effluents. While the tritium seen in the drinking water samples could be attributed to plant effluents, tritium is a radioisotope that occurs naturally in the environment. In 2012, there were no instances where the indicator station results were statistically discernible from the control station results. 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. The REMP trends over the course of time from preoperation to the present are decreasing or have remained fairly constant. This supports the conclusion that there is no adverse radiological impact on the environment or to the public as a result of the operation of Plant Vogtle. 6-1
7.0 ERRATA The following pages are conections to the Vogtle Electric Generating Plant Annual Radiological Environmental Operating Report for 20 it. The conections are a result of the discovery, by Georgia Power Company Environmental Laboratory staff in 2012, of a small positive bias in the 2011 results of OSL environmental dosimeter readings. The method used during 2011 was acceptable at the time but EL dosimetry personnel studied the source of the bias and determined it was based on higher residual dose on the OSL badges as compared to the past Panasonic system. New processing methods are now in place and included in processing procedures. All 2012 environmental OSL processing and reports have included the new methods to remove this small positive bias. The conection has been applied to the 2011 OSL dosimeter results and the conected data are described in the following pages. 7-1
TABLE 3-1 (SHEET lof8) 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 Other Control Pathway Total Detectable Locations Annual Mean Stations (g) Locations Sampled Number of Concentration Mean (b), Mean (b), Mean (b), (Unit of Analyses (MDC) (a) Range Name Distance Mean (b), Range Range Measu remen t) Performed (Fraction) & Direction Range (Fraction) (Fraction) (Fraction) Airborne Gross Beta 10 25.8 Station 15 27.0 24.6 25 .1 Particulates 363 5.4-42.5 Hancock 10.4-42.5 8.1-35.5 3.9-37.8 (fCi/m3) (2591260) Landing Rd. (52152) (52/52) (51151 ) 1.5 miles NW
-oio"1-ma ----- . __
-_. ........._-_ ... _-_....... _.. _.. __... .. ,,- -,. ._-_...__ ._-_._---_.,-_...__._.. - . -.-.--.... --- ~ -- - ---
..-..-.- - ...- ...... -..... -.. .. _ _-_ _-_
....... -_... .... ... ..... _.. _- ......... . ._._..... _..... - _...._...._--_._._...... _.. _.... _... __........
Isotopic 28 Be-7 24 87.1 Station 35 99.3 99.3 85.3
, W,
-..J N 56.9-133.9 Girard 93 .5-104.6 93.5-104.6 82.3-89.6 (20 /20) 6.6 miles SSE (414) (4/4) (4 /4) 1-131 70 NOM NOM NOM NOM Cs-134 50 NOM (c) NOM NOM NOM Cs-137 60 NOM NOM NOM NOM Airborne l-131* 70 63.5 Station 03 76.2 53.2 59.6 Radioiodine 364 30.9-93 .8 Discharge Area 69.2-83.2 24.7-80.4 25.6-76.9 (fCi /m3) (131260) 0.6 miles NE (2152) (4/52) (3 /52) Direct Gamma NA (d) 13 .9 Station 01 18.2 14.0 13.9 Radiation Dose 10.4-19.6 River Bank 16.9-19.6 5.2-20.4 10.2-18.0 (mRl91 days) 158 (64/64) 1.1 mile N (414) (72172) (24/24)
4.3 Direct Radiation In 2011, direct (external) radiation was measured with Landauer lnLight optically stimulated luminescent (OSL) dosimeters which replaced the Panasonic thermoluminescent dosimeters (TLDs). The Panasonic system was retired at the end of 2010 due to the inability to keep the aging badge readers operating reliably. Similar to the TLD protocol of the past, two OSL badges are placed at each station. Each badge contains two elements composed of aluminum oxide crystals with carbon impurity. The gamma dose at each station is based upon the average readings of the elements from the two badges. The two 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 for offsite badges to assure that the badges are on-station and to replace any missing or damaged badges. Two direct radiation 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. The mean and range values presented in the "Other" column in Table 3-1 (page 1 of 8) includes the outer ring stations (stations 17 through 32) as well as stations 35 and 43. As provided in Table 3-1 the average quarterly exposure measured at the indicator stations was 13 .9 mR with a range of 10.4 to 19.6 mR. The average was 0.0 mR less than the average quarterly exposure measured at the control stations (13.9 mR). This difference is not statistically discernible since it is less than the MOO of 1.6 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 2011 ranged from 5.2 to 20.4 mR with an average of 14.0 mR which was 0.1 mR greater than that for the control stations. However, this difference is not discernible since it is less than the MOO of 1.7 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 TLOs from Teledyne to Panasonic. It should be noted however that the differences between indicator and control and outer ring values did not change. 4-1 J 7-3
During 2010, OSL badges were co-located on station with the TLD badges. In 2011, only the OSL badges were placed at each station. Following the change to only OSL badges, the differences between indicator, control, and community locations has been consistent with previous years. An increase noted in 2010 reflects issues (especially during 2n Qtr) with the aging Panasonic TLD reader. The close agreement between the station groups supports the position that the plant is not contributing significantly to direct radiation in the environment. Figure 4.3-1 Average Quarterly Exposure from Direct Radiation 20 18 16 ~ ~ ~ ~~ ;J 14
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Table 4.3-1 Average Quarterly Exposure from Direct Radiation Period Indicator Control Outer Ring (mR) (mR) (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 12.9 13.0 12.9 2002 12.8 12.9 12.6 2003 12.2 12.5 12.4 2004 12.4 12.2 12.3 2005 12.5 13.2 12.9 2006 13.1 12.9 13.0 2007 13.0 12.5 12.7 2008 13.3 13.0 13.1 2009 13.1 13.6 13.3 2010 16.2 16.7 16.6 2011 13.9 13.9 14.0 4-13 7-5
The historical trending of the average quarterly exposures at the special interest areas for the same periods are provided in Fjgure 4.3-2 and listed in Table 4.3-2. These ex posures 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 Average Quarterly Exposure from Direct Radiation at Special Interest Areas 25 I I I I I ! I I I I I I I I 20 . ~A 1 I I I I
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I I I I I I I I I 1 o I Po 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 Year I - Hunting Cabin (Sta 33) - Girard (Sta 35) - . - Rec Center (Sta 43) 4-14 7-6
Tab1e 4.3-2 Average Quarter1y Exposure from Direct Radiation at Special Interest Areas Period Station 33 Station 35 Station 43 (mR) (mR) (mR) Pre-op 16.6 IS.1 IS.3 1987 21.3 18.S IS.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.S 12.0 1993 12.9 13.3 12.1 1994 12.6 13.6 12.0 1995 13.3 13.S 12.3 1996 13.0 13 .6 12.1 1997 13.8 14.4 12.7 1998 13 .S 13.7 12.S 1999 NA 14.S 12.7 2000 NA 14.8 13 .1 2001 NA 14.0 12.6 2002 NA 14.0 12.1 2003 NA 14.1 12.2 2004 NA 14.2 11.7 200S NA IS.2 12.7 2006 NA 14.3 12.6 2007 NA 13.6 11.8 2008 NA 14.4 12.6 2009 NA 14.6 13.0 2010 NA 18.0 16.4 2011 NA IS.6 13.6 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 five deviations from the REMP pertaining to measuring quarterly gamma doses during 2011. 10 second quarter, the dosimeter cases on the badges at Station 47 were chewed by rodents. In third quarter, the dosimeters at Stations 10 and 11 had damaged holding bags. 10 fourth quarter, the dosimeters at Station 8 had moisture in the holding bag, and at Station 2S, the dosimeters were found on the ground at collection time. All of these results passed Chauvenet's Criterion and were retained in the annual OSL dosimeter data set. 4-1S 7-7
The standard deviation for the quarterly result for each Landauer OSL badge was subjected to a self imposed limit of 3.S. Previously with TLDs, this limit had been 1A. However, the OSL readings varied more (between the two elements) than the TLD readings (between the three phosphors). This limit is calculated using a method developed by the American Society for Testing and Materials (ASTM) (ASTM Special Technical Publication lSD, ASTM Manual on Presentation of Data and Control Chart Analysis, Fourth Revision, Philadelphia, PA, October 1976). The calculation is based upon the standard deviations obtained by the EL with the OSL badges during 2010. This limit serves as a flag to initiate an investigation. To be conservative, readings with a standard deviation greater than or equal to 3.S are excluded since the high standard deviation is interpreted as an indication of unacceptable variation in OSL dosimeter response. In 20 II, the OSL results from the following stations were excluded from the data set because their standard deviations were greater than or equal to 3.S: First Quarter: VOlA, Y03B, YOSA, Y07B, Y022B, Y28B, Y31A, Y32A Second Quarter: Y03A, Y13B, Y14B, V22B, YS1B Third Quarter: Y17B, Y32A, YSIB Fourth Quarter: None If one badge at a station exhibited a standard deviation greater than or equal to 3.S, then the reading of the companion badge at each location would be used to determine the quarterly exposure. 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. A major advantage of the OSL badge is that it can be read mUltiple times. A new practice was employed in 20 II to re-read any environmental badges that yielded a standard deviation 2: 3.S. The readings with the lower standard deviation would be reported. 7-8}}