ML20217A654
| ML20217A654 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/31/1997 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9804220304 | |
| Download: ML20217A654 (116) | |
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Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37379-2000 April 16, 1998 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLAN 2 (SQN) - ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT In accordance with Technical Specification 6.9.1.6 for SQN Units 1 and 2, enclosed is the Annual Radiological Environmental Operating Report for 1997.
If you have any questions concerning this matter, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.
Sin el , j Pe r Salas Licensing and Industry Affairs Manager Enclosure cc: See page 2 1
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a 9804220304 971231 PDR ADOCK 05000327
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U.S. Nuclear Regulatory Commission Page 2 April 16, 1998 1
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Mr. R. W. Hernan, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 l
i NRC Resident Inspector
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Sequoyah Nuclear Plant i 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303-3415 l
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ENCLOSURE ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT l
SEQUOYAH NUCLEAR PLANT l
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l ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT SEQUOYAH NUCLEAR PLANT 1997 i
l TENNESSEE VALLEY AUTHORITY ENVIRONMENTAL RADIOLOGICAL MONITORING AND INSTRUMENTATION i i
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_ April 1998
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t TABLE OF CONTENTS l
Table of Contents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii 1
1 l List o f Tables. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv l List o f Figures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v I Executive S ummary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l Introd uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Naturally Occurring and Background Radioactivity. . . . . . . . . . . . . . . . . 2 Electric Power Production. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Site / Plant Description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l Radiological Environmental Monitoring Program. . . . . . . . . . . . . . . . . . . . 9 j Direct Radiation Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Measurement Techniques. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Re s u l t s . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Atmospheric Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 Sample Collection and Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l 16 '
Re s ults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 i Terrestrial Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 18 l Sample Collection and Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 Results.................................................... 19 Aquatic Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 Sample Collection and Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 Res ul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 Assessment and Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 Re s ul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 Conclusions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 Re ferences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 l
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Appendix A Radiological Environmental Monitoring Program and S ampling Locations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 Appendix B 1997 Program Modifications. . . . . . . . . . . . . . . . . . . . . . . . . . . 46 Appendix C Program Deviations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 Appendix D Analytical Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 Appendix E Nominal Lower Limits ofDetection (LLD). . . . . . . . . . . . . . . . 55 Appendix F Quality Assurance / Quality Control Program. . . . . . . . . . . . . . . 61 Appendix G Land Use S urvey. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 Appendix H Data Tables and Figures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 1
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{s LIST OF TABLES
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Table 1 Comparison of Program Lower Limits of Detection with Regulatory Limits for Maximum Annual Average Effluent Concentrations Released to Unrestricted Areas and Reporting Levels. . . . . . . . . . . . . 30 Table 2 Results from the Intercomparison of Environmental Dosimeters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 Table 3 Maximum Dose Due to Radioactive Effluent I R el eas e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 i
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LIST OF FIGURES >
l Figure 1 Tennessee Valley Region. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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Figure 2. Environmental Exposure Pathways of Man Due I to Releases of Radioactive Materials to the l Atmosphere and Lake. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 i
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EXECUTIVE
SUMMARY
t This report describes the radiological environmental monitoring program conducted by TVA in the vicinity of the Sequoyah Nuclear Plant (SQN) in 1997. The program includes the collection i
of samples from the environment and the determination of the concentrations of radioactive materials in the samples. Samples are taken from stations in the general area of the plant and l from areas not influenced by plant operations. Station locations are selected after careful consideration of the weather patterns and projected radiation doses to the various areas around the plant. Monitoring includes the sampling of air, water, milk, foods, vegetation, soil, fish, clams, sediment, and the measurement of direct radiation levels. Results from stations near the plant are compared with concentrations from control stations and with preoperational l measurements to determine potential impacts of plant operations.
l The vast majority of the radioactivity measured in environmental samples from the SQN l program was contributed by naturally occurring radioactive materials or by materials commonly found in the environment as a result of atmospheric nuclear weapons fallout.
l Small amounts of Co-58, Co-60, Cs-134 and Cs-137 were found in sediment samples downstream from the plant. This activity in stream sediment would result in no measurable increase over background in the dose to the general public.
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f-l INTRODUCTION This report describes and summarizes the results of radioactivity measurements made in the vicinity of SQN and laboratory analyses of samples collected in the area. The measurements are made to comply with the requirements of10 CFR 50, Appendix A, Criterion 64 and 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3 and IV.C and to determine potential effects on public health and safety. This report satisfies the annual reporting requirements of SQN Tecimical Specification 6.9.1.6 and Offsite Dose Calculation Manual (ODCM) Administrative Control 5.1.
In addition, estimates of the maximum potential doses to the surrounding population are made L from radioactivity measured both in plant effluents and in environmental samples. The data presented in this report include results from the prescribed program and other useful or interesting information for individuals who do not work with this material routinely.
Naturally Occurrine and Background Radioactivity Most materials in our world today contain trace amounts ofnaturally occurring radioactivity.
Approximately 0.01 percent of all potassium is radioactive potassium-40 (K-40). K-40, with a half-life of 1.3 billion years, is one of the major types of radioactive materials found naturally in our environment. An individual weighing 150 pounds contains about 140 grams of potassium (Reference 1). This is equivalent to approximately 100,000 pCi of K-40 which delivers a dose of 15 to 20 mrem per year to the bone and soft tissue of the body. Naturally occurring radioactive l materials have always been in our environment. Other examples ofnaturally occurring radioactive materials are beryllium (Be)-7, bismuth (Bi)-212 and 214, lead (Pb)-212 and 214, thallium (TI)-208, actinium (Ac)-228, uranium (U)-238 and 235, thorium (Th)-234, radium (Ra)-
226, radon (Rn)-222, carbon (C)-14, and hydrogen (H)-3 (generally called tritium). These naturally occurring radioactive materials are in the soil, our food, our drinking water, and our bodies. The radiation from these materials makes up a part of the low level natural background radiation. The remainder of the natural background radiation comes from outer space. We are all exposed to this natural radiation 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day.
l The average dose equivalent at sea level resulting from radiation from outer space (part of natural i
l background radiation) is about 27 mrem / year. This essentially doubles with each 6600 foot increase in altitude in the lower atmosphere. Another part of natural background radiation comes from naturally occurring radioactive materials in the soil and rocks. Because the quantity of naturally occurring radioactive material varies according to geographical location, the part of the natural background radiation coming from this radioactive material also depends upon the geographical location. Most of the remainder of the natural background radiation comes from the radioactive materials within each individual's body. We absorb these materials from the food we eat which contains naturally occurring radioactive materials from the soil. An example of this is K-40 as described above. Even building materials affect the natural background radiation levels in the environment. Living or working in a building which is largely made of earthen material, such as concrete or brick, will generally result in a higher natural background radiation level than would exist if the same structure were made of wood. This is due to the naturally occurring radioisotopes in the concrete or brick, such as trace amounts of uranium, radium, thorium, etc.
Because the city of Denver, Colorado, is over 5000 feet in altitude and the soil and rocks there contain more radioactive material than the U.S. average, the people of Denver receive around 350 mrem / year total natural background radiation dose equivalent compared to about 295 mrem / year for the national average. People in some locations of the world receive over 1000 mrem / year natural background radiation dose equivalent, primarily because of the greater quantity of radioactive materials in the soil and rocks in those locations.
It is possible to get an idea of the relative hazard of different types of radiation sources by l evaluating the amount of radiation the U.S. population receives from each general type of radiation source. The information in the following table is primarily adapted from References 2 and 3.
1 U.S. GENERAL POPULATION AVERAGE DOSE EQUIVALENT ESTIMATES Source Millirem / Year Per Person Natural background dose equivalent Cosmic 27 Cosmogenic 1 Terrestrial 28 In the body 39 '
- Radon 200 Total 295 l
Release ofradioactive materialin natural gas, mining, ore processing, etc. 5 l Medical (effective dose equivalent) 53 J
i l Nuclear weapons fallout less than 1 l
Nuclear energy 0.28 i Consumer products 0.03 Total 355 (approximately) l As can be seen from the table, natural background radiation dose equivalent to the U.S.
t population normally exceeds that from nuclear plants by several hundred times. This indicates that nuclear plant operations normally result in a population radiation dose equivalent which is insignificant compared to that which results from natural background radiation. It should be noted that the use of radiation and radioactive materials for medical uses has resulted in a similar effective dose equivalent to the U.S. population as that caused by natural background cosmic and terrestrial radiation.
Significant discussion recently has centered around exposures from radon-222 (radon). Radon is an inert gas given off as a result of the decay ofnaturally occurring radium-226 in soil. When dispersed in the atmosphere, radon concentrations are relatively low. However, when the gas is
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trapped in closed spaces, it can build up until concentrations become significant. The National Council of Radiation Protection and Measurements (Reference 2) has estimated that the average annual effective dose equivalent from radon in the United States is approximately 200 mrem / year. This estimated dose is approximately twice the average dose equivalent from all other natural background sources.
l Electric Power Production i
Nuclear power plants are similar in many respects to conventional coal burning (or other fossil I fuel) electric generating plants. The basic process behind electrical power production in both types of plants is that fuel is used to heat water to produce steam which provides the force to turn turbines and generators. However, nuclear plants include many complex systems to control the nuclear fission process and to safeguard against the possibility of reactor malfunction, which could lead to the release of radioactive materials. Very small amounts of these fission and activation products are released into the plant systems. This radioactive material can be transported throughout plant systems and some ofit released to the environment.
l The paths through which radioactivity is released are monitored. Liquid and gaseous effluent monitors record the radiation levels for each release. These monitors also provide alarm mechanisms to prompt termination of any release above limits.
6 Releases are monitored at the onsite points of release and through an environmental monitoring program which measures the environmental radiation in outlying areas around the plant. In this way, not only is the release of radioactive materials from the plant tightly controlled, but measurements are made in surrounding areas to verify that the population is not being exposed to significant levels of radiation or radioactive materials. j l
The SQN ODCM, which is required by the plant Technical Specifications, prescribes limits for the release of radioactive effluents, as well as limits for doses to the general public from the I release of these effluents.
r l The dose to a member of the general public from radioactive materials released to unrestricted l
l areas, as given in NRC guidelines and the ODCM, is limited as follows:
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[ Liauid Effluents Total body 53 mrem / year j Any organ 510 mrem / year Gaseous Effluents Noble gases:
Gamma radiation 510 mrad / year Beta radiation $20 mrad / year
, Particulates:
Any organ $15 mrem / year l The EPA limits for the total dose to the public in the vicinity of a nuclear power plant, established in the Environmental Dose Standard of 40 CFR 190, are as follows:
1 Total body $25 mrem / year Thyroid 575 mrem / year Any other organ 525 mrem / year Appendix B to 10 CFR 20 presents annual average limits for the concentrations of radioactive materials released in gaseous and liquid effluents at the boundary of the unrestricted areas.
Table 1 of this report compares the nominal lower limits ofdetection for the SQN monitoring program with the regulatory limits for maximmn annual average effluent concentrations released to unrestricted areas and levels requiring special reports to the NRC. It should be noted that the levels of radioactive materials measured in the environment are typically below or only slightly above the lower limit of detection. The data presented in this report indicate compliance with the regulation.
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SITE / PLANT DESCRIPTION The SQN is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the westem shore of Chickamauga Lake at Tennessee River Mile (TRM) 484.5.
Figure I shows the site in relation to other TVA projects. The SQN site, containing approximately 525 acres, is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee,14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant (WBN) site.
Population is distributed rather unevenly within 10 miles of the SQN site. Approximately 60 percent of the population is in the general area between 5 and 10 miles from the plant in the sectors ranging from the south, clockwise, to the northwest sector. This concentration is a reflection of suburban Chattanooga and the town of Soddy-Daisy. This area is characterized by considerable vacant land with scattered residential subdivisions. The northem most extent of the urbanization around Chattanooga is approximately 4 miles from the site. The population of Chattanooga is about 160,000, while Soddy-Daisy has approximately 8,500 people. The population within a 10-mile radius of SQN is approximately 75,000.
Residential subdivision growth has continued in the west, north, and east of the plant. There is also some small-scale farming and at least three dairy farms are located within 10 miles of the plant.
1 Chickamauga Reservoir is one of a series of highly controlled multiple-use reservoirs whose primary uses are flood control, navigation, and the generation of electric power. Secondary uses include industrial and public water supply and waste disposal, commercial fishing, and recreation. Public access areas, boat docks, and residential subdivisions have been developed along the reservoir shoreline.
l SQN consists of two pressurized water reactors: each unit is rated at 1183 megawatts (electrical). Fuel was loaded in Unit 1 on March 1,1980, and the unit achieved critically on July 5,1980. Fuel was loaded in Unit 2 in July 1981, and the unit achieved initial criticality on November 5,1981.
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RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Most of the radiation and radioactivity generated in a nuclear power reactor is contained within I the reactor itself or one of the other plant systems. Plant effluent monitors are designed to detect the small amounts released to the environment. Environmental monitoring is a final verification
, that the systems are performing as planned. The monitoring program is designed to check the i
l pathways between the plant and the people in the immediate vicinity and to most efficiently
- l. monitor these pathways. Sample types are chosen so that the potential for detection of l radioactivity in the environment will be maximized. The radiological environmental monitoring program is outlined in Appendix A.
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There are two primary pathways by which radioactivity can move through the environment to humans: air and water (see Figure 2). The air pathway can be separated into two components:
1 the direct (airbome) pathway and the indirect (ground or terrestrial) pathway. The direct airborne l l
pathway consists of direct radiation and inhalation by humans. In the terrestrial pathway, radioactive materials may be deposited on the ground or on plants and subsequently be ingested by animals and/or humans. Human exposure through the liquid pathway may result from drinking water, eating fish, or by direct exposure at the shoreline. The types of samples collected in this program are designed to monitor these pathways.
- A number of factors were considered in determining the locations for collecting environmental samples. The locations for the atmospheric monitoring stations were determined from a critical pathway analysis based on weather patterns, dose projections, population distribution, and land use. Terrestrial sampling stations were selected after reviewing such factors as the locations of
' dairy animals and gardens in conjunction with the air pathway analysis. Liquid pathway stations were selected based on dose projections, water use information, and availability of media such as fish and sediment. Table A-2 (Appendix A, Table 2: This identification system is used for all tables and figures in the appendices.) lists the sampling stations and the types of samples collected. Modifications made to the program in 1997 are described in Appendix B and exceptions to the sampling and analysis schedule are presented in Appendix C.
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l To determine the amount ofradioactivity in the environment prior to the operation of SQN, a l
l preoperational radiological environmental monitoring program was initiated in 1971 and operated until the plant began operation in 1980. Measurements of the same types ofradioactive materials that are measured currently were assessed during the preoperational phase to establish normal background levels for various radionuclides in the environment.
l l The preoperational monitoring program is a very important part of the overall program.
Preoperational knowledge ofpre-existing radionuclide patterns in the environment permits a determination, through comparison and trending analyses, of whether the operation of SQN is impacting the environment and thus the surrounding population.
l The determination ofimpact during the operating phase also considers the presence of control stations that have been established in the environment. Results of environmental samples taken at control stations (far from the plant) are compared with those from indicator stations (near the l plant) to establish the extent of SQN influence.
Samples are analyzed by the Radioanalytical Laboratory of TVA's Environmental Radiological Monitoring and Instrumentation group located at the Western Area Radiological Laboratory (WARL) in Muscle Shoals, Alabama. Analyses are conducted in accordance with written and l approved procedures and are based on accepted methods. A summary of the analysis techniques and methodology is presented in Appendix D. Data tables summarizing the sample analysis l results are presented in Appendix H.
l' l The radiation detection devices used to determine the radionuclide content of samples collected in the environment are generally quite sensitive to small amounts of radioactivity. The sensitivity of the measurcments process is defined in terms of the lower limit of detection (LLD).
A description of the nominal LLDs for the Radioanalytical Laboratory is presented in Appendix E.
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The Radioanalytical Laboratory employs a comprehensive quality assurance / quality control l program to monitor laboratory performance throughout the year. The program is intended to
! detect any problems in the measurement process as soon as possible so they can be corrected. ;
This program includes equipment checks to ensure that the radiation detection instruments are l working properly and the analysis of quality control samples which are included alongside l routine environmental samples. The laboratory participates in the Environmental Protection Agency (EPA) Interlaboratory Comparison Program. In addition, samples split with the State of Tennessee, the State of Alabama and the EPA National Air and Radiation Environmental i
Laboratory provide an independent verification of the overall performance of the laboratory. ' A complete description of the program is presented in Appendix F.
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DIRECT RADIATION MONITORING l
l Direct radiation levels are measured at a number of stations around the plant site. These measurements include contributions from cosmic radiation, radioactivity in the ground, fallout from atmospheric nuclear weapons tests conducted in the past, and any radioactivity that may be present as a results of plant operations. Because of the relatively large variations in background radiation as compared to the small levels from the plant, contributions from the plant may be difficult to distinguish.
Radiation levels measured in the area around the SQN site in 1997 were consistent with levels from previous years and with levels measured at other locations in the region.
Measurement Techniaues Direct radiation measurements are made with thermoluminescent dosimeters (TLDs). When certain materials are exposed to ionizing radiation, many of the electrons which become displaced are trapped in the crystalline structure of the material. They remain trapped for long periods of time as long as the materialis not heated. When heated (thermo), the electrons are released, producing a pulse oflight (luminescence). The intensity of the light pulse is proportional to the amount of radiation to which the material was exposed. Materials which display these characteristics are used in the manufacture of TLDs.
The Panasonic UD-814 dosimeter is ured in the radiological environmental monitoring program for the measurement of direct radiation. This dosimeter contains four elements consisting of one lithium borate and three calcium sulfate phosphors. The calcium sulfate phosphors are shielded 2
by approximately 1000 mg/cm plastic and lead to compensate for the over-response of the detector to low energy radiation.
The TLDs are placed approximately 1 meter above the ground, with two or more TLDs at each station. Sixteen siMions are located around the plant near the site boundary, one station in each
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of the 16 compass sectors. One station is located in each of the 16 compass sectors at a distance of approximately four to five miles from the plant. Dosimeters are also placed at the perimeter and remote air monitoring sites and at 13 additional stations out to approximately 32 miles from the site. The TLDs are exchanged every 3 months and the accumulated exposure on the detectors is read with a Panasonic Model UD-710A sutomatic reader interfaced with a Hewlett Packard Model 9000 computer system. Seven of the locations also have TLDs processed by the NRC.
The results from the NRC measurements are reported in NUREG 0837.
l Since the calcium sulfate phosphor is much more sensitive than the lithium borate, the measured exposure is taken as the median of the results obtained from the calcium sulfate phosphors in the dosimeter badge. The values are corrected for gamma response, system variations, and transit exposure, with individual gamma response calibrations for each element. The system meets or exceeds the performance specifications outlined in Regulatory Guide 4.13 for environmental applications of TLDs.
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Since 1974, TVA has participated in nine of the eleven intercomparisons of environmental I dosimeters conducted by the U.S. Department of Energy and other interested parties. The results, shown in Table 2, demonstrate that direct radiation levels determined by TVA are generally within ten percent of the calculated or known values.
Results Results are normalized to a standard quarter (91.25 days or 2190 hours0.0253 days <br />0.608 hours <br />0.00362 weeks <br />8.33295e-4 months <br />). The stations are grouped according to the distance from the plant. The first group consists of all stations within 1 mile of the plant. The second group lies between I and 2 miles, the third group betwecn 2 and 4 miles, the fourth between 4 and 6 miles, and the fifth group is made up of all stations greater than 6 miles from the plant. Past data have shown that the average results from all groups more than 2 miles from the plant are essentially the same. Therefore, for purposes of this report, all stations 2 miles or less from the plant are identified as "onsite" stations and all others are considered l
"offsite."
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Prior to 1976, direct radiation measurements in the environment were made with dosimeters that were not as precise at lower exposures. Consequently, environmental radiation levels reported in the early years of the preoperational phase of the monitoring program exceed current measurements of background radiation levels. For this reason, data collected prior to 1976 are not included in this report.
The quarterly gamma radiation levels determined from the TLDs deployed around SQN in 1997 are summarized in Table H-1. The results from all measurements at individual stations are presented in Table H-2. The exposures are measured in milliroentgens (mR). For purposes of this report, one milliroetgen, one millirem (mrem) and one millirad (mrad) are assumed to be numerically equivalent. The rounded average annual exposures, as measured in 1997, are shown below. For comparison purposes, the average direct radiation measurements made in the preoperational phase of the monitoring program are also shown.
Annual Average Direct Radiation Levels SQN mR/ Year 1997 1976-79 Onsite Stations 61 79 Offsite Stations 55 63 i
The data in Table H-1 indicate that the average quarterly radiation levels at the SQN onsite stations are approximately 1.5 mR/ quarter higher than levels at the offsite stations. This difference is consistent with levels measured for the preoperation and construction phases of TVA nuclear power plant sites where the average levels onsite were generally 2-6 mR/ quarter higher than levels offsite. The causes of these differences have not been isolated; however, it is postulated that the differences are attributable to combinations ofinfluences such as natural variations in environmental radiation levels, earth-moving activities onsite, and the mass of
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concrete employed in the constmetion of the plant. Other undetennined influences may also play a part.
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i Figure H-1 compares plots of the data from the onsite or site boundary stations with those from the offsite stations over the period from 1976 through 1997.
The results reported in 1997 are consistent with direct radiation levels identified at locations which are not influenced by the operation of SQN. There is no indication that SQN activities increased the background radiation levels normally observed in the areas surrounding the plant.
ATMOSPHERIC MONITORING The atmospheric monitoring network is divided into three groups identified as local, perimeter, and remote. Four local air monitoring stations are located on or adjacent to the plant site in the general directions of greatest wind frequency. Four perimeter air monitoring stations are located in communities out to about 10 miles from the plant, and four remote air monitors are located out to approximately 20 miles. The monitoring program and the locations of monitoring stations are identified in the tables and figures of Appendix A. The remote stations are used as control or baseline stations.
Results from the analysis of samples in the atmospheric pathway are presented in Tables H-3 and H-4. Radioactivity levels identified in this reporting period are consistent with natural background and radionuclides produced as a result of fallout from previous nuclear weapons tests. There is no indication of an increase in atmospheric radioactivity as a result of SQN.
Samnie Collection and Analysis Air particulates are collected by continuously sampling air at a flow rate of approximately 2 cubic feet per minute (cfm) through a 2-inch glass fiber filter. The sampling system consists of a pump, magnehelic gauge for measuring the drop in pressure across the system, and a dry gas i
meter. This allows an accurate determination of the volume of air passmg through the filter. l This system is housed in a building approximately 2 feet by 3 fet t by 4 feet. The filter is contained in a sampling head mounted on the outside of the monitur building. The filter is replaced every 7 days. Each filter is analyzed for gross beta activity about 3 days after collection to allow time for the radon daughters to decay. Every 4 weeks composites of the filters from each location are analyzed by gamma spectroscopy.
Gaseous radioiodine is collected using a commercially available cartridge containing TEDA impregnated charcoal. This system is designed to allect iodine in both the elemental form and as organic compounds. The cartridge is located in the same sampling head as the air particulate filter and is downstream of the particulate filter. The cartridge is changed at the same time as the r
particulate filter and samples the same volume of air. Each cartridge is analyzed for I-131 by a complete gamma spectroscopy analysis.
Rainwater is sampled by use of a collection tray attached to the monitor building. The collection l tray is protected from debris by a screen cover. As water drains from the tray, it is collected in l one of two 5-gallon containers inside the monitor building. A 1-gallon sample is removed from the container every 4 weeks. Any excess water is discarded. Rainwater samples are held to be analyzed only if the air particulate samples indicate the presence of elevated activity levels or if fallout is expected. For example, rainwater samples were analyzed during the period of fallout l following the accident at Chemobyl in 1986. Since no plant related air activity was detected in environmental samples in 1997, no rainwater samples from SQN were analyzed in this reporting period.
1 Results The results from the analysis of air paniculate samples are summarized in Table H-3. Gross beta activity in 1997 was consistent with levels reported in previous years. The average level was 0.021 pCi/m' for both indicator and control locations. The annual average of the gross beta l activity in air particulate filters at these stations for the years 1971-1997 are presented in Figure l l
H-2. Increased levels due to fallout from atmospheric nuclear weapons testing are evident, especially in 1971,1977,1978, and 1981. Evidence of a smallincrease resulting from the l l
Chernobyl accident can also be seen in 1986. These patterns are consistent with data from l monitoring programs conducted during the preoperation and construction phases at TVA nuclear plant sites.
1 Only natural radioactive materials were identified by the monthly gamma spectral analysis of the air particulate samples. No fission or activation products were detected. As shown in Table H-4, l I-131 was not detected in any of the charcoal canister samples collected in 1997.
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TERRESTRIAL MONITORING l I l l 1
! Terrestrial monitoring is accomplished by collecting samples of environmental media that may transport radioactive material from the atmosphere to humans. For example, radioactive material may be deposited on a vegetable garden and be ingested along with the vegetables or it may be I deposited on pasture grass where dairy cattle are grazing. When the cow ingests the radioactive material, some ofit may be transferred to the milk and consumed by humans who drink the milk.
Therefore, samples of milk, vegetation, soil, and food crops are collected and analyzed to i
determine potential impacts from exposure through this pathway. The results from the analysis j
\
of these samples are shown in Tables H-5 through H-13.
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A land use survey is conducted annually to locate milk producing animals and gardens within a 5-mile radius of the plant. Three dairy farms are located on the east side of the river between 4 l
and 6 miles from the plant. Two farms with at least one milk producing animal have been identified within 2 miles of the plant. The locations with the highest calculated dose to individuals drinking the milk are included in the sampling program. The dairy located about 5 l
miles northeast of the plant and the two farms near the plant are considered indicator stations. !
The results of the 1997 land use survey are presented in Appendix G.
Samole Collection and Analysis Milk samples are collected every 2 weeks from the three indicator locations and from at least one of three control dairies. These samples are placed on ice for transport to the Radioanalytical Laboratory. A specific analysis for I-131 and a gamma spectroscopy analysis are performed on each sample and Sr-89,90 analysis is performed quarterly.
Vegetation is being sampled every 4 weeks from one farm that had milk producing animals in the l past. An additional sample is collected from one control station. The samples are collected by cutting or breaking enough vegetation to provide between 100 and 200 grams of sample. Care is taken not to include any soil with the vegetation. The sample is placed in a container with 1650 ml of 0.5 N NaOH for transport back to the Radioanalytical Laboratory for I-131 analysis. A second sample of between 750 and 1000 grams is also collected from each location. After drying i
and grinding, these samples are analyzed by gamma spectroscopy. Once each quarter, the samples are ashed after the gamma analysis is completed and analyzed for Sr-89,90.
1
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Soil samples are collected annually from the air monitoring locations. The samples are collected with either a " cookie cutter" or an auger type sampler. After drying and grinding, the sample is analyzed by gamma spectroscopy. When the gamma analysis is complete, the sample is ashed and analyzed for Sr-89,90. i i
Samples representative of food crops raised in the area near the plant are obtained from I individual gardens, comer markets, or cooperatives. Types of foods may vary from year to year as a result of changes in the local vegetable gardens. In 1997 samples of apples, cabbage, corn, green beans, potatoes, and tomatoes were collected from local vegetable gardens. The edible portion of each sample is analyzed by gamma spectroscopy.
Results The results from the analysis of milk samples are presented in Table H-5. No radioactivity which could be attributed to SQN was identified. All I-131 results were less than the established nominal LLD of 0.4 pCi/ liter. Strontium-90 was detected in approximately one third of the samples. The Sr-90 levels are consistent with historical data reported in milk as a result of fallout from atmospheric nuclear weapons tests (Reference 1). Figure H-3 displays the average Sr-90 concentrations measured in milk since 1971. The concentrations have steadily decreased as a result of the 28-year half-life of Sr-90 and the washout and transport of the element through the soil over the period. The average Sr-90 concentration reported from indicator stations in 1997 was 6.39 pCi/ liter. The average concentration reported from control stations was 2.09 pCi/ liter. By far the predominant isotope reported in milk samples was the naturally occurring K-40. An average of approximately 1350 pCi/ liter of K-40 was identified in all milk samples.
I As has been noted in this report for previous years, the levels of Sr-90 in milk samples from I
small farms producing milk for private consumption have been consistently higher than the levels found in milk from commercial dairy farms. Samples of feed and water supplied to the animals were analyzed in 1979 in an effort to determine the source of the strontium. Analysis of dried hay samples indicated levels of Sr-90 slightly higher than those encountered in routine vegetation samples. Analysis ofpond water indicated no significant strontium activity.
This phenomenon was observed during the preoperational radiological monitoring near SQN at farms where only one or two cows were being milked for private consumption of the milk. It is postulated that the feeding practices of these small farms differ from those of the larger dairy farmers to the extent that fallout from atmospheric nuclear weapons testing may be more concentrated in these instances. Similarly, Hansen, et al. (Reference 4), reported an inverse
- relationship between the levels of Sr-90 in milk and the quality of fertilization and land management. These phenomenon account for the slightly higher levels of Sr-90 measured in milk samples from the two farms near the SQN site used as indicator locations.
i Results from the analysis of vegetation samples (Table H-6) were similar to those reported for milk. AllI-131 values were less than the nominal LLD. Strontium-90 was identified in a total of seven samples at concentrations ranging from 15.0 to 46.6 pCi/Kg. These concentrations are consistent with results produced by nuclear weapons fallout. Again, the largest concentrations identified were for the naturally occurring isotopes K-40 and Be-7.
l A total of thirteen soil samples were collected and analyzed. The soil samples contained measurable levels of Cs-137 with the maximum concentration being 0.91 pCi/g. One sample contained 0.50 pCi/gm Sr-90. These concentrations are consistent with levels previously reported from fallout. All other radionuclides reported were naturally occurring isotopes (Table H-7).
A plot of the annual average Cs-137 concentrations in soil is presented in Figure H-4. Like the levels of Sr-90 in milk, concentrations of Cs-137 in soil are steadily decreasing as a result of the t
cessation of weapons testing in the atmosphere, the 30-year half-life of Cs-137 and transport through the environment.
Radionuclides reported in food samples were naturally occurring. The maximum K-40 value was 3440 pCi/kg in potatoes. Analysis of these samples indicated no contribution from plant activities. The results are reported in Tables H-8 through H-13.
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AQUATIC MONITORING Potential exposures from the liquid pathway can occur from drinking water, ingestion of edible fish and invertebrates, or from direct radiation exposure from radioactive materials deposited in the river sediment. The aquatic monitoring program includes the collection of samples of surface water, groundwater, drinking water supplies, fish, Asiatic clams (there is no known human consumption of these clams from the Tennessee River), and bottom and shoreline sediment.
Samples from the reservoir are collected both upstream and downstream from the plant. I Results from the analysis of aquatic samples are presented in Tables H-14 through H-22.
Radioactivity levels in water, fish, and clams were consistent with background and/or fallout levels previously reported. The presence of Co-58, Co-60, Cs-134 and Cs-137 was identified in some bottom sediment samples and Cs-137 was detected in shoreline sediment. There is no direct exposure pathway to the public through radioactivity in bottom sediment. The projected exposure to the public through the shoreline sediment would be less than 0.1 mrem / year.
Samole Collection and Analysis Samples of surface water are collected from the Tennessee River downstream and upstream of j
the plant using automatic sampling systems. A timer turns on the system at least once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the sample is collected into a compositejug. A 1-gallon sample is removed from the compositejug at 4-week intervals and the remaining water in the jug is discarded. The I composite sample is analyzed for gamma emitting radionuclides and for gross beta activity. A quarterly composite sample is analyzed for Sr-89,90 and tritium. A change in the surface water I sampling that eliminated a second downstream location was implemented during 1997. The change is discussed in Appendix B.
Samples are collected by an automatic sampling system at the first downstream drinking water intake and at the water intake for the city of Dayton located approximately 19 miles upstream.
These samples are collected in the same manner as the surface water samples and analyzed by gamma spectroscopy and for gross beta activity. At other selected locations, grab samples are collected from drinking water systems which use the Tennessee River as their source. These samples are analyzed every 4 weeks by gamma spectroscopy and for gross beta activity. A quarterly composite sample from each station is analyzed for Sr-89,90 and tritium.
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The sample collected by thegutomatic sampling device is taken directly from the river at the intake structure. Since the sample at this point is raw water, not water processed through the water treatment plant, the control sample should also be unprocessed water. Therefore, the upstream surface water sample is also considered as a control sample for drinking water.
Groundwater is sampled from an onsite well and from a private well in an area unaffected by SQN. The quarterly composite samples are prepared for each location and analyzed by gamma i
spectroscopy. Analyses are also performed for gross beta activity, Sr-89,90 and tritium.
Samples of commercial and game fish species are collected semiannually from each of two reservoirs: the reservoir on which the plant is located (Chickamauga Reservoir) and the upstream reservoir (Watts Bar Reservoir). The samples are collected using a combination of netting techniques and electrofishing. Samples of all species are prepared from filleted fish.
After drying and grinding, the samples are analyzed by gamma spectroscopy.
Bottom and shoreline sediment are collected semiannually from selected TRM locations using a dredging apparatus or divers. The samples are dried and ground and analyzed by gamma spectroscopy.
Samples of Asiatic clams are collected semiannually from one location below the plant and one location above the plant. There is no known use of these clams for human consumption. The clams are usually collected in the dredging or diving process with the sediment. Enough clams are collected to produce approximately 50 grams of wet flesh. The flesh is separated from the shells, and the dried flesh samples are analyzed by gamma spectroscopy.
i Results j
There were no fission or activation product radionuclides identified from the gamma i spectroscopy or specific analyses performed on surface water samples. Gross beta activity above the nominal LLD value was measured in most surface water samples. Concentrations in downstream samples averaged 2.9 pCi/ liter while the upstream samples averaged 3.1 pCi/ liter.
}
The values were consistent with previously reported levels. A trend plot of the gross beta activity in surface water samples from 1971 through 1997 is presented in Figure H-5. A summary table of the results is shown in Table H-14.
There were no fission or activation product radionuclides identified in drinking water samples.
Average gross beta activity was 2.6 pCi/ liter for the downstream stations and 3.2 pCi/ liter at the control stations. The results are shown in Table H-15 and a trend plot of the gross beta activity .
1 in drinking water from 1971 to the present is presented in Figure H-6. I No fission or activation products were detected by the gamma spectroscopy or specific analyses I performed on well water. Gross beta concentrations in samples from the onsite well averaged 2.4 pCi/ liter, while the average from the offsite well was 8.4 pCi/ liter. The results are presented in Table H-16.
Cesium-137 was identified in a total of five fish samples. The maximum concentration measured in samples from indicator locations was 0.07 pCi/g, while the maximum for control samples was 0.05 pCi/g. Plots of the annual Cs-137 concentrations in the filleted samples are presented in Figures H-7, H-8, and H-9. Since the concentrations downstream are essentially equivalent to the upstream levels, the concentrations of Cs-137 are most likely the result of fallout or other upstream effluents rather than activities at SQN. Other radioisotopes found in fish were naturally I
occurring with the most notable being K-40. The concentrations of K-40 ranged from 8.1 pCi/g to 17.3 pCi/g. The results are summarized in Tables H-17, H-18, H-19. l l
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Radionuclides of the types produced by nuclear power plant operations were identified in bottom sediment samples. The materials identified were Cs-137, Cs-134, Co-60, and Co-58. The average Cs-137 concentration measured for samples from the downstream locations was 0.56 pCi/g and the average concentration for control locations was 0.77 pCi/g. The presence of Cs-137 was measured in three samples from downstream shoreline sediment monitoring locations.
The maximum concentration was 0.07 pCi/g. Samples of shoreline sediment from the control location did not contain Cs-137 above the LLD value. The concentrations of Cs-137 in shoreline sediment are consistent with previously identified fallout levels and are most likely not a result of SQN operations. The Co-60 concentrations measured in downstream bottom sediment samples averaged 0.24 pCi/g. Co-60 was not identified in upstream samples. I Cs-134 was identified in one downe.m c'. ream bottom sediment sample. The concentration reponed was 0.05 pCi/g. Co-58 was identified in two downstream samples at an average f
concentration of 0.06 pCi/g and Co-58 was measured in one upstream sample at a concentration of 0.20 pCi/g. A dose assessment of the impact to the general public from this activity produces a negligible dose equivalent. Results from the analysis of bottom sediment samples are shown in Table H-20. Co-58, Co-60, and Cs-134 were not identified in shoreline sediment. Results from the analysis of shoreline sediment samples are shown in Table H-21.
Graphs of the Cs-137 and Co-60 concentrations in bottom sediment are presented in Figures H-10 and H-11, respectively. Figure H-12 presents a plot of the Cs-137 concentrations measured in shoreline sediment since 1980.
Only naturally occurring radioisotopes were identified in clam flesh samples. The results from the analysis of these samples are presented in Table H-22.
l ASSESSMENT AND EVALUATION l J
Potential doses to the public are estimated from measured effluents using computer models. 4 These models were developed by TVA and are based on methodology provided by the NRC in Regulatory Guide 1.109 for determining the potential dose to individuals and populations living in the vicinity of a nuclear power plant. The doses calculated are a representation of the dose to a
" maximum exposed individual." Some of the factors used in these calculations (such as ingestion rates) are maximum expected values which will tend to overestimate the dose to this
" hypothetical" person. In reality, the expected dose to actual individuals is significantly lower.
The area around the plant is analyzed to determine the pathways through which the public may receive an exposure. As indicated in Figure 2, the two major ways by which radioactivity is introduced into the environment are through liquid and gaseous effluents.
For liquid effluents, the public can be exposed to radiation from three sources: drinking water from the Tennessee River, eating fish caught in the Tennessee River, and direct exposure to radioactive material due to activities on the banks of the river (recreational activities). Data used to determine these doses are based on guidance given by the NRC for maximum ingestion rates, exposure times, and distribution of the material in the river. Whenever possible, data used in the dose calculation are based on specific conditions for the SQN area.
For gaseous effluents, the public can be exposed to radiation from several sources: direct i radiation from the radioactivity in the air, direct radiation from radioactivity deposited on the ground, inhalation of radioactivity in the air, ingestion of vegetation which contains radioactivity deposited from the atmosphere, and ingestion of milk from animals which consumed vegetation containing deposited radioactivity. The concentrations of radioactivity in the air and the soil are l
estimated by computer models which use the actual meteorological conditions to determine the distribution of the effluents in the atmosphere. Again, as many of the parameters as possible are based on actual site specific data.
Results The estimated doses to the maximum exposed individual due to radioactivity released from SQN in 1997 are presented in Table 3. These estimates were made using the concentrations of the
~
liquids and gases measured in the effluent monitoring points. Also shown are the regulatory limits for these doses and a comparison between the calculated dose and the corresponding limit.
The maximum calculated whole body dose equivalent from measured liquid effluents as presented in Table 3 is 0.022 mrem / year, or 0.7 percent of the limit. The maximum organ dose equivalent from gaseous effluents is 0.058 mrem / year. This represents 0.39 percent of the NRC limit. A more complete description of the effluents released from SQN and the corresponding doses projected from thue effluents can be found in the SQN Annual Radioactive Effluent Release Report.
As stated earlier in this report, the estimated increase in radiation dose equivalent to the general public resulting from the operation of SQN is negligible when compared to the dose from natural background radiation. The results from each environmental sample are compared with the concentrations from the corresponding control stations and appropriate preoperational and background data to determine influences from the plant. During this report period, Co-60, Co-58, Cs-134, and Cs-137 were detected in sediment. The Cs-137 concentrations measured in sediment is consistent with fallout levels identified in samples both upstream and downstream from the plant. The Co-60, Co-58, and Cs-134 identified in sediment samples downstream from the plant would produce no measurable increase in the dose to the general public. No increases of radioactivity attributable to SQN have been seen in water samples.
Dose estimates were made from concentrations of radioactivity found in samples of environmental media. Media evaluated included, but are not limited to, air, milk, food products, drinking water, fish, soil and shoreline sediment. Inhalation, ingestion and direct doses estimated for persons at the indicator locations were essentially identical to those determined for persons at control stations. More than 99 percent of those doses were contributed by the naturally occurring radionuclide K-40 and by Sr-90 and Cs-137, which are long-lived radioisotopes found in fallout .
1 4
from nuclear weapons testing. Concentrations of Sr-90 and Cs-137 are consistent with levels measured in TVA's preoperational radiological environmental monitoring programs. Figures H-3 and H-4 and Figure H-8 through H-10 indicate that concentrations of Sr-90 and Cs-137 in the environment have decreased since the cessation of atmospheric weapons testing in 1981. This l
decrease is the result of the decay of the two isotopes and the redistribution of the materials in the i
environment.
1 Conclusions It is concluded from the above analysis of the environmental sampling results and from the trend plots presented in Appendix H that the exposure to members of the general public which may have been attributable to SQN is negligible. The radioactivity reported herein is primarily the result of fallout or natural background radiation. Any activity which may be present as a result of plant operations does not represent a significant contribution to the exposures ofMembers of l the Public.
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REFERENCES l
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- 1. Merril Eisenbud, Environmental Radioactivity. Academic Press, Inc., New York, NY,1987.
1
- 2. National Council on Radiation Protection and Measurements, Report No. 93, " Ionizing Radiation Exposure of the Population of the United States," September 1987.
- 3. United States Nuclear Regulatory Commission, Regulatory Guide 8.29," Instruction Concerning Risks from Occupational Radiation Exposure," July 1981.
i
- 4. Hansen, W.G., Campbell, J. E., Fooks, J. H., Mitchell, H.C., and Eller C.H., Farming Practices and Concentrations of Emission Products in Milk. U.S. Department of Health, Education, and Welfare; Public Health Service Publication No. 999-R-6, May 1964.
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l Table 1 COMPARISON OF PROGRAM LOWER LIMITS OF DETECTION WITH THE REGULATORY LIMITS FOR MAXIMUM ANNUAL AVERAGE EFFLUENT CONCENTRATIONS RELEASED TO UNRESTRICTED AREAS AND REPORTING LEVELS Concentrations in Water. DCi/ Liter Concentrations in Air. oCi/ Cubic Meter Effluent Reporting Lower limit Effluent Reporting Lowerlimit Concentration' Level 2 of Detection 8 Concentration' Level 2 of Detection' H3 1,000,000 20,000 300 100,000 Cr-51 500,000 45 30,000 0.02 Mn 54 30,000 1,000 5 1,000 0.005 Co-58 20,000 1,000 5 1,000 0.005 Co-60 30,000 300 5 50 0.005 1 Zn-65 5,000 300 10 400 0.005 St89 8,000 5 1,000 0.0011 Sr90 500 2 6 0.0004 Nb-95 30,000 400 5 2,000 0.005 4 Zr-95 20,000 400 10 400 0.005 Ru-103 30,000 5 900 0.005 Ru-106 3,000 40 20 0.02 I-131 1,000 2 0.4 200 0.9 0.03 Cs-134 900 30 5 200 )
10 0.005 j Cs 137 1,000 50 5 200 20 0.005 Ce 144 3,000 30 40 0.011 Ba-140 8,000 200 25 2,000 0.015 La-140 9,000 200 10 2,000 0.01 Note: 1 pCi = 3.7 x10-2 Bq.
Note: For those reporting levels that are blank, no value is given in the reference.
I Source: Table 2 of Appendix B to 10 CFR 20.1001-20.2401 2 Source: SQN Offsite Dose Calculation Manual, Table 2.3 2 3 Source: Table E-1 of this report. I Table 2 Results from the Intercomparison of Environmental Dosimeters i
Calculated Average, all Exposure % Difference % Difference TVA Results Respondents (See Note 1) TVA: Respondents:
Ys.E HERIH HESHI HEI.nl Calculated Calculated Field Dosimeters l 74 15.0 16.3 16.3 -8.0 0.0 77 30.4 31.5 34.9 -12.9 9.7 79 13.8 16.0 14.1 2.1 13.5 81 31.8 30.2 30.0 6.0 0.7 82 43.2 45.0 43.5 -0.7 3.4 84 73.0 75.1 75.8 3.7 -0.9 86a 33.2 28.9 29.7 11.8 -2.7 86b 9.4 10.1 10.4 -9.6 -2.9 93a 24.4 26.4 27.0 -9.6 -2.2 93b 27.6 26.4 27.0 2.2 -2.2 96a 16.9 18.9 19.0 -10.9 -0.5 96b 17.6 18.9 19.0 7.4 -0.5 Low Irradiated Dosimeters 74 27.9 28.5 30.0 -7.0 -5.0 79 12.1 12.1 12.2 -0.8 -0.8 86 I8.2 16.2 17.2 5.8 -5.8 93a 24.9 25.0 25.9 -3.9 -3.5 93b 27.8 25.0 25.9 7.3 -3.5 High Irradiated Dosimeters 77 99.4 86.2 91.7 8.4 -6.0 79 46.1 43.9 45.8 0.7 -4.1 81a 84.1 75.8 75.2 11.8 0.8 81b 102.0 90.7 88.4 15.4 2.6 82a 179.0 191.0 202.0 -11.4 -5.4 82b 136.0 149.0 158.0 -13.9 -5.7 84a 85.6 77.9 79.9 7.1 2.5 84b 76.8 73.0 75.0 2.4 -2.7 93a 67.8 69.8 72.7 -6.7 -4.0 93b 80.2 69.8 72.7 10.3 -4.0 96a 60.7 55.2 58.1 4.5 -5.0 96b 59.4 55.2 58.1 2.2 -5.0 Notes: 1. The calculated exposure is the "known" exposure determined by the testing agency.
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Table 3 Maximum Dose Due to Radioactive Emuent Releases Sequoyah Nuclear Plant 1997 mrem / year Dose From Liquid Emuents 1997 NRC Percent of LEt D2K Liulit NRC Limit
. Total Body 0.022 3 0.7 Any Organ 0.037 10 0.4 Doses From Gaseous Emuents 1997 NRC Percent of LDS D2R LiHlit NRC Limit Noble Gas 0.049 10 0.49 (Gamma)
Noble Gas 0.132 20 0.66 (Beta)-
l Any Organ 0.058 15 0.39 t
l Total Cumulative Dose 1997 EPA Percent of bgg Dose LiUlit EPA Limit Total Body or Any Other Organ 1.49E-01 25 <l .0 Thyroid 1.22E-01 75 <l.0 i
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l APPENDIX A RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM AND SAMPLING LOCATIONS I
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l Table A-2 i
SEQUOYAH NUCLEAR PLANT l RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS l
Map Approximate indicator (I)
Location Distance or Samples Numbera Station Sector (Miles) Control (C) Collectedb 2 LM-2 N 0.8 i AP,CF,R,S 3 LM-3 SSW 2.0 1 AP,CF,R,S i 4 LM-4 NE I.5 I AP,CF,R,S !
5 LM-5 NNE 1.8 i AP,CF.R.S l 7 PM 2 SW 3.8 I AP,CF,R,S 8 PM-3 W 5.6 I AP,CF,R,S 9 PM-8 SSW 8.7 I AP,CF,R,5 10 PM-9 WSW 2.6 I AP,CF,R,S II RM-I SW 16.7 C AP,CF.R.S 12 RM 2 Nr4E 17.8 C AP,CF,R.S 13 RM 3 ESE 11.3 C AP,CF,R,S 14 RM-4 WNW 18.9 C AP,CF,R,S 15 Farm B NE 43.0 C M 16 Farm C NE 16.0 C M 17 Farm S NNE 12.0 C M,V 18 Farm J WNW l.1 ! M 19 Farm HW NW l .2 1 M,WC 20 Farm EM N 2.6 I V 24 Well No. 6 NNE 0.15 I W 31 TRM 473.0 - 10.7d g pW (C. F. Industries) 32 TRM 469.9 -
13.8d I pw (E.1. DuPont) 33 TRM 465.3 -
18.4d I pw (Chattanooga) 34 TRM 497.0 - 13.3d C SWe 35 TRM 503.8 -
20.id C PW (Dayton) 36 TRM 496.5 -
6.7d I ss 44 TRM 478.0 -
5.7d g ss 46 Chickamauga Reservoir (TRM 471530) -
I/C F.CL 47 Watts Bar Reservoir (TRM 530-602) -
C F 48 Farm H NE 4.2 I M l
- a. See Figures A-1, A-2, and A-3 I
- b. Sample codes: l AP = Air particulate filter PW = Public Water SS = Shoreline Sediment CF = Charcoal filter R = Rainwater SW = Surface water CL = Clams S = Soil V = Vegetation !
F = Fish SD = Sediment W = Well water l M = Milk l
- c. A control for well water.
i d. Distance from plant discharge (TRM 483.7).
( c. Surface water sample also used as a control for public water.
i
Table A 3 SEQUOYAH NUCLEAR PLANT THERMOLUMINESCENT DOSIMETER (TLD) II) CATIONS Map Approximate Location Onsite (Onf Distance or W 3g[ igg SSW IC jgg3pg tr!P]$1) Offsite (Ofn SSW 2.0 On 4 NEIA NE 1.5 On 5 NNE1 NNE 1.8 On 7 SW-2 SW 3.8 Off 8 W3 W 5.6 04 9 SSW-3 SSW 8.7 Off 10 WSW 2A WSW 2.6 Off 11 SW3 SW 16.7 Off 12 NNE-4 NNE 17.8 Off 13 ESE-3 ESE 11.3 Off 14 WNW-3 WNW 18.9 Off 49 N1 N 0.6 On 50 N-2 N 2.1 Off 51 N-3 N 5.2 oft 52 N4 N 10.0 oft 53 NNE-2 NNE 4.5 Off 54 NNE-3 NNE 12.1 Off 55 NE-1 NE 2.4 Off 56 NE-2 NE 4.1 Off 57 ENE-1 ENE 0.4 On 58 ENE-2 ENE 5.1 oft 59 E-1 E I .2 On 60 E-2 E 5.2 Off 61 ESE-A ESE 0.3 On 62 ESE1 ESE I.2 On 63 ESE-2 ESE 4.9 Off 64 SE-A SE 0.4 On 65 E-A E 0.3 On 66 SE-l SE I.4 On 67 SE-2 SE 1.9 On 68 SE-4 SE 5.2 Off 69 SSE-l SSE I .6 On 70 SSE-2 SSE 4.6 Off 71 S-1 S 1.5 On 72 S-2 S 4.7 Off 73 SSW-1 SSW 0.6 On 74 SSW 2 SSW 4.0 Off 75 SW 1 SW 0.9 On 76 WSW-1 WSW 0.9 On 77 WSW-2 WSW 2.5 Off 78 WSW 3 WSW 5.7 Off 79 WSW 4 WSW 7.8 Off 80 WSW-5 WSW 10.1 Off 81 W1 W 0.8 On 82 W-2 W 4.3 oft 83 WNW-1 WNW 0.4 On 54 WNW-2 WNW 5.3 Off 85 NW.! NW 0.4 On 86 NW2 NW 5.2 Oli 87 NNW1 NNW 0.6 On 88 NNW 2 NNW l.7 On 89 NNW-3 NNW 5.3 oft 90 SSW 1B SSW l .5 On
- a. See Figures A-l A-2,and A-3.
- b. TLDs designated "onsite" are located 2 miles or less from the plant;"ofTsite"are located more than 2 miles from the plant.
Figure A-1 Radiological Environmental Mcnitoring Locations Within 1 mile of the Plant 348.75 N 11.25 NNW NNE 328.25 33.75 NW 2 NE 303.75 \ 49 f 58.25 WNW , ENE 8*) 281.25 s N
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73 / q@ SW v# gsSY SE ! 213.75 1 148.25 i s i .25 f 188.75 Scale O Mile 1 j Figure A-2 Radiological Environmental Monitoring Locations Between 1 and 5 miles from the Plant 348.75 N' 11.25 NNW NNE 326.25 i 33.75 NW [ NE 303.75 56.25 56 WNW / Y 5 55 ENE 281.25 gps,s 78.75 82 1 edp,e W- 1 i 0# -E 59 10 2 258.75 77 0 [6 101.25 9 71 69 WSW [ 3 ESE s 236.25 go 123.75 74 SW SE
'46 e70 213.75 7,2 146.25 SSW SSE 1s t.25 s 1sa.75 8CALE 6 1 2 MILES l
l Figure A-3 Radiological Environmental Monitoring Locations l More than 5 miles from the Plant 348.75 " .A. 11.26
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l i l ! i APPENDIX B 1997 PROGRAM MODIFICATIONS I [ I i 1 { i l l l
L Appendix B Radiological Environmental Monitoring Program Modification l In March 1997, modifications were made to implement program changes resulting from revisions to the SQN ODCM. These changes involved reductions in the number of downstream surface I water and sediment sampling locations. The second downstream automatic surface water sampler located at TRM 473.2 was eliminated. A decision was made that this was not necessary since it was only about 3 miles upstream from the automatic sampler located at TRM 469.9. During the same program change, the sediment sampling conducted at TRM 472.8 was also discontinued. This was the third downstream sampling point for bottom sediment and was almost eleven miles from the plant discharge. Based on historical data the detennination was
- made that the two sampling points closer to the plant discharge provided adequate monitoring for this environmental media.
These changes are summarized in Table B-1. l
Table B-1 Radiolonical Environmental Monitorine Prograrn Modifications Dats Station Location Remarks l March 97 TRM 473.2 10.5 miles (d Discontinued the collection of surface water. March 97 TRM 472.8 10.9 miles (d Discontinued the collection of bottom sediment. (d Miles from plant discharge (TRM 483.7) l l l I i APPENDIX C PROGRAM DEVIATIONS i
}
l l I
-49 I
L
Appendix C Radiological Environmental Monitoring Program Deviations During 1997, there were three sampling periods when the air particulate filter and charcoal cartridge could not be collected from one of the twelve sampling locations due to equipment
- problems. The effected stations and dates are listed in the Table C-1. In each case the problem was corrected and a sample was collected as scheduled the next week.
The ground water sample scheduled for collection from the on site monitoring well was not l available on July 9,1997 due to problems with the electrical power service to the sampling station. The problem was corrected and a sample was collected at the next scheduled collection period. A total of four milk samples were missed during the year because milk was not available at the farm or dairy. These missed samples resulted in deviations from the scheduled program but did not represent a problem of noncompliance with the ODCM required program. Table C-1 provides additional l details on the missed samples. 1 l l l
-50
Table C-1 Radiolonical Environmental Monitorinn Pronram Deviations
}
Dalt S.latigg Location Remarks l l 01/28/97 RM-4 18.9 miles WNW Air particulate filter and charcoal cartridge were not collected due to equipment problems. The drive belt on the sampling pump was broken. l l 04/15/97 PM-3 5.6 miles W Air particulate filter and charcoal cartridge were not collected due to equipment problems. The electric power l was off at the sampling station. 1 07/09/97 Well No. 6 0.15 miles NNE Adequate sample volume was not available in the sampler for the monthly sample. Here was a temporary disruption in the electric power supply for the sampler. A monthly l sample is collected from this sampler and held for use in a ; l quarterly composite. For the effected quarterly sampling l period, the composite was prepared from the other samples i collected during the quarter. 09/03/97 Farm C 16.0 miles NE The milk was picked up early by the processor and a sample was not available when the sample collector arrived at the farm. His is one of three controllocations. Samples were collected at the other control locations during this sampling period. I1/12/97 Farm J 1.1 miles WNW This location is a small farm with only one cow. The cow 11/24/97 was not producing milk during this period and no sample 12/09/97 was available. l 12/09/97 PM-2 3.8 miles SW Air particulate filter and charcoal cartridges were not collected due to equipment problems. %e drive belt on the sampling pump was broken. l 1 l 1 APPENDIX D ANALYTICAL PROCEDURES i l l l l
l l Appendix D Analytical Procedures ! l Analyses of environmental samples are performed by the radioanalytical laboratory located at the i Western Area Radiological Laboratory facility in Muscle Shoals, Alabama. All analysis procedures are based on accepted methods. A summary of the analysis techniques and methodology follows. l The gross beta measurements are made with an automatic low background counting system. Normal counting times are 50 minutes. Water samples are prepared by evaporating 500 ml of samples to near dryness, transferring to a stainless steel planchet and completing the evaporation process. Air particulate filters are counted directly in a shallow planchet. The specific analysis ofI-131 in milk, water, or vegetation samples is perfonned by first l isolating and purifying the iodine by radiochemical separation and then counting the final l precipitate on a beta-gamma coincidence counting system. The normal count time is 50 minutes. l With the beta gamma coincidence counting system, background counts are virtually eliminated and extremely low levels of activity can be detected. ! l After a radiochemical separation, samples analyzed for Sr-89, 90 are counted on a low : background beta counting system. The sample is counted a second time after a 7-day ingrowth period. From the two counts the Sr-89 and Sr-90 concentrations can be determined. i Water samples are analyzed for tritium content by first distilling a portion of the sample and then counting by liquid scintillation. A commercially available scintillation cocktail is used. l Gamma analyses are performed in various counting geometries depending on the sample type and volume. All gamma counts are obtained with germanium type detectors interfaced with a ( l l
computer based multichannel analyzer system. Spectral data reduction is performed by the computer program HYPERMET. t l i The charcoal cartridges used to sample gaseous radiciodine are analyzed by gamma spectroscopy l using a high resolution gamma spectroscopy system with germanium detectors. l l The necessary efficiency values, weight-efficiency curves, and geometry tables are established and maintained on each detector and counting system. A series ofdaily and periodic quality j control checks are performed to monitor counting instrumentation. System logbooks and control l l l charts are used to document the results of the quality control checks. ' l , l l { 1 i 1 1 l l l l 1 l 1 i 4 I l \ l l l l l APPENDIX E 1 NOMINAL LOWER LIMITS OF DETECTION (LLD) 1 ! l i ( i r Appendix E l Nominal Lower Limits of Detection i Sensitive radiation detection devices can produce a signal even when no radioactivity is present in a sample being analyzed. This signal may come from trace amounts ofradioactivity in the components of the device, from cosmic rays, from naturally occurring radon gas, or from electronic noise. The signal registered when no activity is present in the sample is called the background. f The point at which the signal is determined to represent radioactivity in the sample is called the l I critical level. This point is based on statistical analysis of the background readings from any particular device. However, any sample measured over and over in the same device will give different readings, some higher than others. The sample should have a well-defined average reading, but any individual reading may vary from that average. In order to determine the activity present in a sample that will produce a reading above the critical level, additional j statistical analysis of the background readings is required. The hypothetical activity calculated l from this analysis is called the lower limit of detection (LLD). A listing of typical LLD values g that a laboratory publishes is a guide to the sensitivity of the analytical measurements performed by the laboratory. Every time an activity is calculated from a sample, the background must be subtracted from the j sample signal. For the very low levels encountered in environmental monitoring, the sample signals are often very close to the background. The measuring equipment is being used at the l lirnit ofits capability. For a sample with no measurable activity, which often happens, about half the time its signal should fall below the average machine background and half the time it should be above the background. If a signal above the background is present, the calculated activity is l compared to the calculated LLD to determine if there is really activity present or if the number is ! I an artifact of the way radioactivity is measured. 1 j i l t
1 A number of factors influence the LLD, including sample size, count time, counting efliciency, chemical processes, radioactive decay factors, and interfering isotopes encountered in the sample. The most likely values for these factors have been evaluated for the various analyses I performed in the environmental monitoring program. The nominal LLDs calculated from these values, in accordance with the methodology prescribed in the ODCM, are presented in Table E-1. i The maximum values for the lower limits of detection specified in the ODCM are shown in Table E-2. { The nominal LLDs are also presented in the data tables. For analyses for which LLDs have not been established, an LLD of zero is assumed in determining if a measured activity is greater than 1 the nominal LLD. 1 t nir e od l
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Table E-2 Maximum Values for the Lower Limits of Detection (LLD) Specified by the SQN Offsite Dose Calculation Manual Airborne Particulate Food Water or Gases Fish Milk Products Sediment Analysis If.k'L DCi/m'_ _ h wet oCi/L __ oC /kg. wet pCi/ka, dry gross beta 4 1 x 102 N.A. N.A. N.A. N.A. H-3 2000* N.A. N.A. N.A. N.A. N.A. l Mn 15 N.A. 130 N.A. N.A. N.A. l Fe-59 30 N.A. 260 N.A. N.A. N.A. Co-58,60 15 N.A. 130 N.A. N.A. N.A. Zn-65 30 N.A. 260 N.A. N.A. N.A. Zr-95 30 N.A. N.A. N.A. N.A. N.A. Nb-95 15 N.A. N.A. N.A. N.A. N.A. 1131 16 7 x 10-2 N.A. I 60 N.A. Cs-134 15 5 x10-2 130 15 60 150 Cs-137 18 6 x 10 2 150 18 80 180 Ba-140 60 N.A. N.A. 60 N.A. N.A. La-140 15 N.A. N.A. 15 N.A. N.A.
- a. If no dnnking water pathway exists, a value of 3000 pCi/ liter may be used.
- b. If no dnnkwg water pathway exists, a value of 15 pCi/ liter may be used.
l l 1 l
1 i 1 I 1 APPENDIX F QUALITY ASSURANCE / QUALITY CONTROL PROGRAM l i i l 1 l l Appendix F Ouality Assurance /Ouality Control Program l A thorough quality assurance program is employed by the laboratory to ensure that the environmental monitoring data are reliable. This program includes the use of written, approved procedures in performing the work, a complete training and retraining system, internal self assessments of program performance, audits by various extemal organizations, and a laboratory l l quality control program. j
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The quality control program employed by the radioanalytical laboratory is designed to ensure l that the sampling and analysis process is working as intended. The program includes equipment checks and the analysis ofquality control samples along with routine samples. I Radiation detection devices can be tested in a number of ways. There are two primary tests l which are performed on all devices. In the first type, the device is operated without a sample on the detector to determine the background count rate. The background caunts are usually low values and are due to machine noise, cosmic rays, and/or trace amour of radioactivity in the materials used to construct the detector. Charts of background counts are kept and monitored to ensure that no unusually high or low values are encountered. In the second test, the device is operated with a known amount of radioactivity present. The 1 number of counts registered from such a radioactive standard should be very reproducible. These { reproducibility checks are also monitored to ensure that they are neither higher nor lower than expected. When counts from either test fall outside the expected range, the device is inspected l for malfunction or contamination. It is not returned to service until it is operating properly. I In addition to these two general checks, other quality control checks are performed on the l variety of detectors used in the laboratory. The exact nature of these checks depends on the type of device and the method it uses to detect radiation or store the information obtained.
-62
Quality control samples of a variety of types are used by the laboratory to verify the performance of different portions of the analytical process. These quality control samples may be blanks, replicate samples, blind samples, or cross-checks. Blanks are samples which contain no measurable radioactivity or no activity of the type being measured. Such samples are analyzed to determine whether there is any contamination of equipment or commercial laboratory chemicals, cross-contamination in the chemical process, or l interference from isotopes other than the one being measured. i l Duplicate samples are generated at random by the sample computer program which schedules the I collection of the routine samples. For example, if the routine program calls for four milk samples every week, on a random basis each farm might provide an additional sample several times a year. These duplicate samples are analyzed along with other routine samples. They provide information about the variability ofradioactive content in the various sample media. i i ! If enough sample is available for a particular analysis, the laboratory staff can split it into two portions. Such a sample can provide information about the variability of the analytical process since two identical portions of material are analyzed side by side. l Analytical knowns are another category of quality control sample. A known amount of radioactivity is added to a sample medium. The lab staff know the radioactive content of the l sample. Whenever possible, the analytical knowns contain the same amount of radioactivity each time they are run. In this way, analytical knowns provide immediate data on the quality of the measurement process. A portion of these samples are also blanks. Blind spikes are samples containing radioactivity which are introduced into the analysis process lisguised as ordinary environmental samples. The lab staff does not know the sample contains radioactivity. Since the bulk of the ordinary workload of the environmental laboratory contains no m asurable activity or only naturally occurring radioisotopes, blind spikes can be used to test the detection capability of the laboratory or can be used to test the data review process. If an analysis routinely generates numerous zeroes for a particular isotope, the presence of the isotope is brought to the attention of the laboratory supervisor in the daily review process. Blind spikes test this process since the blind spikes contain radioactivity at levels high enough to be detected. Furthermore, the activity can be put into such samples at the extreme limit of detection (near the LLD) to determine whether or not the laboratory can find any unusual radioactivity whatsoever. At present,5 percent of the laboratory workload is in the category ofintemal cross-checks. These samples have a known amount of radioactivity added and are presented to the lab staff labeled as cross-check samples. This means that the quality control staff knows the radioactive content or "right answer" but the lab personnel performing the analysis do not. Such samples test the best performance of the laboratory by determining if the lab can find the "right answer". These samples provide information about the accuracy of the measurement process. Further information is available about the variability of the process if multiple analyses are requested on the same sample. Like blind spikes or analytical knowns, these samples can also be spiked with low levels of activity to test detection limits. 4 A series of cross-checks is produced by the EPA in Las Vegas. These interlaboratory comparison samples or " EPA cross-checks" are considered to be the primary indicator of laboratory performance. They provide and independent check of the entire measurement process that cannot be easily provided by the laboratory itself. That is, unlike internal cross-checks, EPA l cross-checks test the calibration of the laboratory detection devices since different radioactive standards produced outside TVA are used in the cross-checks. The results of the analysis of these samples are reported back to EPA which then issues a report of all the results of all participants. These reports indicate how well the laboratory is doing compared to others across the nation. Like intemal cross-checks, the EPA cross-checks provide information to the laboratory about the precision and accuracy of the radioanalytical work it does. f a l l The results of TVA's participation in the EPA Interlaboratory Comparison Program are presented in Table F-1. For 1997, all EPA cross -check sample concentrations measured by TVA's laboratory were within i 3-sigma of the EPA reported values. 1 TVA splits certain environmental samples with laboratories operated by the States of Alabama ! j and Tennessee and the EPA National Air and Radiation Environmental Laboratory in Montgomery, Alabama. When radioactivity has been present in the environment in measurable quantities, such as following atmospheric nuclear weapons testing, following the Chernobyl i incident, or as naturally occurring radionuclides, the split samples have provided TVA with another level ofinformation about laboratory performance. These samples demonstrate performance on actual environmental sample matrices rather than on the constructed matrices used in cross-check programs. The quality control data are routinely collected, examined and reported to laboratory supervisory personnel. They are checked for trends, problem areas, or other indications that a portion of the analytical process needs correction or improvement. The end results is a measurement process i that provides reliable and verifiable data and is sensitive enough to measure the presence of .. radioactivity far below the levels which could be harmful to humans. )
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l l l APPENDIX G 1 l l l LAND USE SURVEY 1
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I i 1 1
L Appendix G I Land Use Survey l A land use survey is conducted annually to identify the location of the nearest milk producing animal, the nearest residence, and the nearest garden of greater than 500 square feet producing fresh leafy vegetables in each of 16 meteorological sectors within a distance of 5 miles from the plant. The land use survey is conducted between April 1 and October 1 using appropriate techniques such as door-to-door survey, mail survey, telephone survey, aerial survey, or information from l local agricultural authorities or other reliable sources. l In order to identify the locations around SQN which have the greatest relative potential for i impact by the plant, radiation doses are projected for individuals living near SQN. These projections use the data obtained in the survey and historical meteorological data. They also assume that releases are equivalent to the design basis source terms. The calculated doses are relative in nature and do not reflect actual exposures received by individuals living near SQN. Calculated doses to individuals based on measured effluents from the plant are well below applicable dose limits (see Assessment and Evaluation Section and Table 3). I In response to the 1997 SQN land use survey, annual doses were calculated for air submersion, vegetable ingestion, and milk ingestion. External doses due to radioactivity in air (air l submersion) are calculated for the nearest resident in each sector, while doses from drinking milk or eating foods produced near the plant are calculated for the areas with milk producing animals and gardens, respectively. l t
Air submersion doses were calculated for the same locations as in 1996. There were no changes in the resulting value since there were no changes in the nearest resident distance. Doses calculated for ingestion of home-grown foods changed slightly for two sectors due to small changes in the location of the nearest garden. Also no garden was found in the ENE sector. I For milk ingestion, projected doses were consistent with those calculated for 1996 except for small variances due to changes in the feeding factor value. This factor was updated for each location based on the data obtained during the land use survey. Samples are being taken from the three farms with the highest projected doses and the highest X/Q values. The 1997 survey identified the same potential milk producing location that was identified in the 1996 survey at approximately 1.5 miles NNW. There were milk goats at this location at the time of the survey but the animals were not currently producing milk. Periodic contact with the resident at this I location has indicated that milk was still not being produced as ofMarch 1998. No sampling is currently required or conducted at this location. Tables G-1, G-2, and G-3 show the comparative relative calculated doses for 1996 and 1997. 1 Table G 1 SEQUOYAli NUCLEAR PLANT Relative Projected Annual Air Submersion Dose to the Nearest Resident Within Five Miles of Plant mrem / year 1996 Survey 1997 Survey Approximate Approximate Distance Annual Distance Annual Smtat Milsa Dnas Miles D9ss N 0.8 0.12 0.8 0.12 NNE 1.5 0.07 1.5 0.07 NE 1.5 0.06 1.5 0.06 ENE 1.3 0.02 1.3 0.02 E 1.0 0.02 1.0 0.02 ESE 1.0 0.02 1.0 0.02 SE 1.1 0.02 1.1 0.02 SSE 1.3 0.03 1.3 0.03 S 1.2 0.10 1.2 0.10 SSW l.3 0.15 1.3 0.15 SW l.4 0.06 1.4 0.06 WSW 0.6 0.05 0.6 0.05 W 0.6 0.06 0.6 0.06 WNW l.1 0.02 1.1 0.02 NW 0.8 0.04 0.8 0.04 NNW 0.5 0.14 0.5 0.14 I
Table G-2 SEQUOYAH NUCLEAR PLANT Relative Projected Annual Dose to Child's Critical Organ from Ingestion of Home-Grown Foods mrem / year ! 1996 Survey 1997 Survey i Approximate Approximate Distance Annual Distance Annual Sgstor hi!Les Dose Mileg Dgg N 1.1 2.25 1.1 2.25 NNE 1.6 2.10 1.6 ' 2.10 NE 2.1 1.18 2.7 0.78 ENE 1.6 0.61 (a) ) E 2.0 0.28 2.0 0.28 j ESE 1.3 0.04 ' 1.3 0.40 SE 1.1 0.74 2.0 0.30 SSE 1.3 1.00 1.3 1.00 S 1.4 2.45 1.4 2.45 SSW l.7 3.50 1.7 3.50 SW 2.4 1.02 2.4 1.02 WSW 0.7 1.32 0.7 1.32 W l.2 0.63 1.2 0.63 WNW l.1 0.62 1.1 0.62 NW 0.9 1.16 0.9 1.16 NNW 0.5 4.26 0.5 4.26 (a) Garden not found within 5 miles. i l l Table G-3 SEQUOYAH NUCLEAR PLANT Relative Projecced Annual Dose to Receptor Thyroid from Ingestion of Milk mrem / year Approximate Distance Annual Dose XQ Location Esp. tor (MilesP 1996 1997 sl,!p', Farm H NE 4.7 0.048 0.043 2.94 E-7 Farm HS6d E 4.6 0.010 0.010 6.74 E-8 Farm JH6d ESE 3.9 0.004 0.004 6.79 E-8 i Farm y WNW l.1 0.029 0.040 3.99 E-7 Farm HW- NW l.2 0.057 0.040 5.48 E-7 I
- a. Distances measured to nearest property line,
- b. Grade A dairy.
- c. Milk sampled at this location.
! d. Not currently sampled in the SQN monitoring program. t
1 i l l l J l I APPENDIX H l DATA TABLES AND FIGURES 1 l l l l i l i i
Table H - 1 DIRECT RADIATION LEVELS Average External Gamma Radiation Levels at Various Distances from Sequoyah Nuclear Plant for Each Quarter - 1997 mR / Quarter (a) Distance per annum Mues l Average Extemal Gamma Radiation Levels (b) l mR/yr 1st qtr 2nd qtr 3rd qtr 4th qtr 0-1 15.3 i 1.6 16.5 i 1.8 17.3 i 1.7 16.6 i 1.6 66 1-2 12.9
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- 2.1 14.412.1 13.4 i 1.7 54 4-6 12.9 i 1.3 14.111.5 14.511.6 13.911.5 55
>6 12.7 i 1.2 13.8 i 1.4 14.4 i 1.3 13.9 i 1.3 55 Average, 0 - 2 miles 14.2 i 2.0 15.312.1 16.1 i 2.1 15.312.1 61 (onsite)
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> 2 miles 12.711.4 13.9 i 1.6 14.4 i 1.6 13.811.5 55 (offsite)
(a) Field periods normalized to one standard quarter (2190 hours) ! (b) Average of the individual measurements in the set i 1 standard deviation of the set
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