ML19296A781

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Radiological Assessment of Individual Dose Resulting from Routine Operation-Demonstration of Compliance w/40CFR190
ML19296A781
Person / Time
Site: Westinghouse
Issue date: 01/28/1980
From:
NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML19296A779 List:
References
NUDOCS 8002190032
Download: ML19296A781 (15)
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Text

{{#Wiki_filter:_ DOCKET NO: 70-1151 LICENSEE: Westinghouse Electric Corporation FACILITY: Fuel Fabrication Plant Columbia, South Carolina

SUBJECT:

RADIOLOGICAL ASSESSMENT OF INDIVIDUAL DOSE RESULTING FROM ROUTINE OPERATION - OEMONSTRATION OF COMPLIANCE WITH 40 CFR 190 I.

Background

The EPA uranium fuel cycle standard, as specified in 40 CFR 190,1 litaits the total dose to an individual from radioactivity associated with the routine operation of nuclear fuel cycle facilities to 25 mrem /yr to the total body, 75 mrem /yr to the thyroid, and 25 mrem /yr to any other organ. The standard will become effective on December 1,1979, for all uranium fuel fabrication plants used for the production of LWR fuel. The Westinghouse Company's (the licensee) plant is an existing uranium oxide fuel facility and is subject to the EPA , standards. Based on the most current plant operation, emission and monitoring data, the NRC staff conducted the following radiological assessment to determine if the licensee meets the EPA's standard on fuel cycle facilities. As a result of this assessment, an action level on the effluent release rate from routine operation of the facility will be established to provide reasonable assurance that the licensee complies with the standards for continued operation. II. Discussion A. Descriotion of the Facility 1. Plant Ooeration - General Westinghouse owns and operates a facility at Columbia, South Carolina, for the fabrication of uranium dioxide fuel assemblies for light water moderated power reactors. Following the current building expansion, the plant capacity is expected to be less than 1200 MTV/ year. 2. Chemical Process The process of chemically converting the uranium fluoride raw material to uranium dioxide is carried out in a system of parallel process lines, most of which use the conventional wet ammonium diuranate (ADU) process. There is, however, some conversion using the proprietary, direct conversion fluidized bed (DCFB) process. 1

2 For either the ADV or DCFB process, UFs is received at a maximum enrichment of 4.15% U-235 in standard two-and-one-half-ton cylinders in appropriate NRC and 00T approved shipping packages. As required, a UFs cylinder is removed from the UFs cylinder storage area and connected to one of the conversion lines. The UFs is vaporized by heating the cylinder in one of the steam chambers located in the UFs vaporization area adjacent to the conversion lines. In the ADV process, the vaporized UFs is hydrolyzed to uranyl fluoride (UO F ) and HF by mixing with water. The uranyl 22 fluoride is subsequently converted to an ADV slurry by addition of ammonium hydroxide solution. The ADU slurry is dewatered by centrifugation and the ADU sludge fed to an externally fired, cylindrical calciner. In the calciner, the ADV is converted to the solid UO2 product by heat and the introduction of hydrogen. The ammonia and steam in the calciner off gases are scrubbed by a water scrubber and the gases then passed through a high efficiency particulate air (HEPA) filter assembly before discharge to the atmosphere. The dry uranium oxide powder is conveyed from the calciner through a milling operation and into storage containers which are sampled, closed, and identified. In the DCFB process, a continuous flow procedure, the UFs and other reactant gases enter series-connected electrically-heated fluidized bed reactors. Excess steam, hydrogen, and gaseous reaction products (HF) are removed from the bulk of the solid 002 produced in the upper part of the final reactor and exit to the off gas system. Discharge of UO2 powder from the lower part of the reactor is governed by bed weight controllers. The uranium oxide product is received continuously in a holdup vessel and then pneumatically transported to the uranium oxide milling system. Milling, packaging, and storage of the uranium oxide is similar to the corresponding operations in the ADU lines. An off gas scrubber system condenses HF and excess steam from the exhaust gases, which are HEPA filtered before release through a continuously sampled stack. 3. Mechanical process The U02 powder is blended, mechanically treated and pressed into pellets. The pellets are sintered in a reducing atmosphere, ground to a standard diameter and welded into Zircaloy' tubes. The loaded fuel rods are assembled into bundles, tested, and packaged for shipment. 4. Auxiliary Process Ooerations The auxiliary process operations include (1) a scrap recovery process that involves chemical dissolution and solvent extraction purification of recovered uranium, and (2) incineration of combustible waste.

3 B. Waste Confinement and Effluent Controls 1. Gaseous Effluents Operations involving the use of radioactive materials in unsealed physical forms are limited to low-enrichment (< 4.15 wt% U-235) uranium in the fuel manufacturing facilities or associated analytical laboratory. The ventilation systems installed in these facilities are designed so that all of the air from zones used to handle or process uranium is treated to remove essentially all the uranium prior to its release to the atmosphere. The effluents are released from a number of short stacks and roof vents, as shown in Figure 1. Nonradioactive process gases are also vented through several short stacks and vents, also shown in Figure 1. Each monitored release stack is equipped with a device that continuously draws a sample through a low porosity filter. The filter paper is then removed periodically and analyzed for uranium. 2. Liouid Effluents The aqueous process waste solution, primarily filtrate from the ADV process lines, is circulated through filters before being pumped to tanks in the Waste Treatment Facility. The main constituents of the process liquid wastes are ammonium fluoride and uranium. Through the addition of lime and caustic, the fluoride is converted to insoluble CaF, which is removed by 2 centrifugation or settling in a series of holding lagoons. (The ultimate disposal of the CaF 2 from the lagoons is addressed in the plant decommissioning plan.) The settling lagoons also remove uranium and some ammonia. Most of the ammonia is recovered by distillation and returned to the ADU process following pH adjustment with caustic. Site sanitary sewage is treated in an extended aeration package plant and discharged into a biological oxidation / settling, polishing lagoon. The lagoon effluent is then chlorinated and mixed with treated liquid process waste at the facility lift station. The combined waste is passed through a final aerator, followed by pH adjustment as required, and pumped to the Congaree River. 3. Solid Waste Disposal Noncombustible solid waste that is or could be uranium contaminated, but where uranium recovery efforts are not justified, is collected in boxes for ultimate disposal at a Government-licensed waste disposal site.

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5 C. Semiannual Effluent Emission Data Section 70.59 of 10 CFR Part 70 requires that the licensee submit effluent monitoring reports on a semiannual basis. Tables 1 and 2 summarize the results of the airborne and liquid radioactive effluent measurements for the period July 1975 through December 1978. D. Descriotion of the Site Environment Related to Radiological Assessment at the Maximum Nearest Resicent The following description of the site environment provides information specific to the assessment of the impact on individuals from radiological effluents released during normal operation of the facility. A more complete description of the site environment may be found in the Environmental Report 2 issued by the licensee in March 1975. 1. Site Location The Nuclear Fuel Columbia Site (NFCS) is located in central South Carolina in Richland County. It is approximately 8 miles Table 1 Semiannual Airborne Releases (pCi) Period U-234 U-235 U-238 July-Dec 1975 963.6 50.3 244.1 Jan-June 1976 1128.1 56.8 270.7 July-Dec 1976 1095.0 118.7 288.4 Jan-June 1977 1198.7 44.3 285.9 July-Dec 1977 1916.0 90.5 440.5 Jan-June 1978 1186.3 48.6 284.1 July-Oec 1978 919.3 40.0 182.7 Table 2 Semiannual Liouid Releases (uCi) Period U-234 U-235 U-238 July-Dec 1975 52,322.4 2,732.2 13,251.4 Jan-June 1976 35,968.5 1,810.0 8,632.5 July-Dec 1976 40,941.0 4,446.0 10,812.0 Jan-June 1977 46,664.0 1,735.0 11,126.0 July-Dec 1977 63,100.0 2,994.0 14,474.0 Jan-June 1978 49,036.0 2,023.0 11,763.0 July-Dec 1978 78,095.0 3,409.0 15,529.0

6 southeast of the Colun ia city limits and is reached by South Carolina Highway 48. Nearoy towns, industrial plants, public facilities, the Congaree River, and transportation links are shown in Figure 2. The site is bounded by Route 48 to the north, the Vestal Lumber Manufacturing Company property to the east, the Liberty Life Insurance Company property to the south, and the Burrel Manning property to the west. The manufacturing facilities lie in the center of approximately 1,158 acres. The fuel fabrication facilities, holding ponds, parking lot, and landscaped grounds occupy approximately 60 acres. Figure 3 snows the plant boundary and adjacent properties. Figure 3 also presents the site elevations. The plant floor is 142 feet above mean sea level (MSL). Plant-site drainage-flow follows original drainage patterns to Sunset Lake, Mill Creek, and the Congaree River. The area around the site is primarily flat to the north and flat and swampy in other directions. Most of the unused portions of the site (approximately 1,098 acres) will be left in its natural state. 2. Land and Water Use Approximately 70 percent of the land within 5 miles of the NFCS is forest or forested swampland. Agricultural land occupies approximately 20 percent of the study area, most of which is in the northern and eastern portions. Livestock production in the study area is not especially significant. Egg production is noted to be important locally as evidenced by the presence of the Columbia Egg Division of FCX, Inc., in Hopkins. Beef cattle were observed at a number of farms, though not in any great quantity. At present, there are only four active dairy farms in the entire county, two of which are in the study area. One dairy, with a 700 cow capacity, is located 2.2 miles northwest of the NFCS on the north side of Route 48. Other livestock, such as broilers, hogs and sheep, were not observed. Less than one percent of the land within the study area is used for industrial purposes. With the exception of the Carolina Eastman plant that lies 4.75 miles directly west of the NFCS (38% of the business work force) and the Square D Company that lies 4.66 miles north-northeast (46% of the business work force), the facilities within the study area are considered light industrial -- warehousing, shipping, packaging, metal fabrication, etc. - and generally employ less than 40 persons each.

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9 The major stream near the plant is the Congaree River. Principal tributaries to the Congaree include Gills Creek at Columbia, Mill Creek near the NFCS plant, and Beaver Creek near Hopkins. Sunset Lake is a shallow, artificial impoundment of Mill Creek within the NFCS boundary. There are presently no routine releases to Sunset Lake and environmental monitoring of the lake water has not shown radioactive contamination in detectable levels. The Congaree River is characterized by high levels of suspended solids, sandy bottoms, and sand beaches. At the NFCS discharge point, the Congaree is about 500 feet wide and 9 feet (or less) deep. The flow of the Congaree River at Columbia averages 9,166 cubic feet per second. 3. Diffusion Climatolocy Annual average joint frequency distributions of wind speed and direction were obtained from onsite meteorological data collected at the NFCS from August 1, 1972, through July 31, 1973, using the STAR program.3 The results are presented in the licensee's Environmental Report. For annual atmospheric dispersion calculations, the above joint frequency distributions were used. The resulting dispersion factors, X/Q, were calculated using the Gaussian Plume model and diffusion coefficients for Pasquill-type turbulence.4 Since the NFCS effluent release points are generally lower than two and one-half times the height of adjacent solid structures, the release was conservatively assumed to occur at ground level with credit granted for building wake effects.5 The annual average X/Q as a function of distance up to 50 miles from the site were calculated for 16 sectors as shown in Table 3. 4. The Nearest Residence The licensee's Environmental Report describes the location of the nearest offsite residence as 3,300 feet northeast of the center of the manufacturing building. The nearest site boundary is 1,800 feet to the north-northwest. Individual dose commitments were calculated for both an individual at the lccation of the nearest resident and conservatively, at the nearest site boundary. E. Environmental Imoact from Routine Plant Ooeration 1. Methodology for Radiological Assessment The general approach which will be used to demonstrate compliance with the dose limits of 40 CFR 190 is as follows: (i) Effluents released from plant operation will be monitored to determine the quantities of radionuclides discharged into the environment.

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11 (ii) Environmental dose models developed by NRC to estimate dose commitment rates from all significant pathways will be used. It is only when noncompliance is suspected that a detailed environmental monitoring program will be required to supplement such effluent monitoring. The above approaches to demonstrate compliance are in conformance with the recommendations of the EPA as specified in their Final Environmental Statement (FES) for Environmental Radiation Protection Requirements for Normal Operations of Activities in the Uranium Fuel Cycle.6 The source terms (radioactive effluent release rates) from the NFCS plant operations are actual measured values. The atmospheric dispersion model is based on Reg. Guide 1.111.7 Other environmental pathways and models are based on Reg. Guide 1.1098 with the exception that for the inhalation pathway, the dose conversion factors for various organs were generated using the ICRP Task Group Lung Models72 The dose conversion factors from the Task Group Lung Model depend on particle size and solubility of the radioactive compounds released. If this information is not available from the licensee, a reasonable or conservative approach will be applied for the radiological impact assessment. For example, the particle size was assumed to have an average diameter (AMAD) of 0.3 um for effluents passing through HEPA filters and 1.0 um AMAD for particles not passing through HEPA filters; the particles released were conservatively assumed to be 100% in an insoluble form which will provide a maximum lung dose for the inhalation calculation and then 100% in a soluble form which will provide a maximum bone dose for the ingestion calculation. It is only when such conservative assumptions are critical to the standards (i.e., near or exceeding 25 mrem /yr) that the licensee is required to conduct studies to obtain information for a more realistic assessment. 2. Maximum Individual Dose The radiological impacts were assessed by calculating the maximum total body and organ doses to the closest resident, who is living at about two-thirds of a mile northeast of the NFCS plant. Except where specified, the term " dose," as referred to in this assessment, is actually a 50 year dose commitment, that is, the total dose to the reference organ from one year's intake of radionuclides which will accrue during the remaining lifetime (50 years) of the individual. Tables 1 and 2 summarize the semiannual release rates of radiological airborne effluents which were used as source terms for the radiological assessment. The release rates are measured values. The respect.ee semiannual release rates were averaged for a representative six-month rate and that value doubled for the annual release value used in the calculations.

12 For liquid effluents discharged into the Congaree River, it was conservatively assumed that the uranium is in a soluble form. It was further assumed that the liquid release is only diluted by the river flow at the discharge. For airborne effluents released into the environment, the pathways considered for the individual dose calculations included (a) direct irradiation from either ground or shoreline deposition, (b) direct inhalation, and (c) ingestion pathways (vegetation, meat, milk) due to airborne deposition. For liquid effluent releases, the pathways included (a) potable water, (b) aquatic food (fish), and (c) shoreline deposition. The models and various assumptions used to calculate doses from the above environmental pathways can be referred to in greater detail in Regulatory Guide 1.109. Table 4 summarizes the results of the calculated maximum annual dose from airborne and liquid effluents to the nearest resident. As shown in Table 4, the critical pathway is due to inhalation resulting in a maximum dose to the lung of 2.96 mrem /yr. The above calculations assume a normal adult; the staff also considered a critical individual at the nearest residence. The critical individual in the inhalation pathway is an infant (0-1 years of age). The lung dose to the infant will be increased by a factor of about 1.8, i.e., 5. 33 mrem /yria less than 25% of the environmental standard. Therefore, the staff concluded that the maximum annual lung dose is well below the 25 mrem annual limit as specified in 40 CFR Part 190 and that there is no adverse effect due to the release of effluents from the NFCS's normal plant operation. F. Conclusion and Recommendation The normal operation of the NFCS fuel fabrication plant results in the release of a minute quantity of radioactivity into the environment. Based on past operation, the annual release of radioactivity used for these calculations was approximately 3100 pCi of uranium in airborne effluents and 134.53 mci of uranium in liquid effluents. The nearest resident is located at about two thirds of a mile northeast of the plant site. The annual lung dose to the critical individual at the nearest residence was estimated under conservative assumptions to be 5.33 mrem /yr which represents only 21% of the 25 mrem limit of the EPA standard as specified in 40 CFR 190. The staff, therefore, concludes there is no adverse impact from the release of radioactivity due to routine operation of the NFC3 fuel fabrication plant. The staff recognizes that the nearest resident located two thirds of a mile northeast of the facility might not represent the potential maximum impact from the NFCS operation. The staff estimated that the maximum impact in the unrestricted area could be at the nearest site boundary, 1800 feet to the north-northwest of the center of the manufacturing buildings. The X/Q at this location is 1.54 x 10 s 3 sec/m. This land is presently owned by the Burrel Mining Company.

13 Table 4 Estimated Maximum Annual Dose from Airborne and Liquid effluents to the Nearest Resloent Pathways Organ Oose (millirem /yr) Total-Body Lung Bone A. Air Effluents 1. Direct Irradiation 1.43 x 10 5 2. Direct Inhalation" 1.54 x 10 3 2.96 2.50 x 10 2 3. Ingestion due to Airborne Deposition a. Vegetation"* 6.06x10[3 9.81 x 10 2 b. Meat 2.90 x 10 s 4.73 x 10 3 4 c. Milk 1.20 x 10 4 1.95 x 10 8. Licuid Effluents 1. Potable Water 6.14 x 10 4 1.00 x 10 2 2. Aquatic Food (Fish) 3.49x10[5 5.71 x 10 4 3. Shoreline Deposit 5.45 x 10 to Total (millirem /yr) 8.41 x 10 3 2.96 1.36 x 10 1 ~ Assume 80% residence time. Includes nonleafy and leafy vegetables. Since site specific information was not available, the staff assumed 76% of the produce and 100% of the leafy vegetables ingested were grown in a garden at the nearest residence.

14 If a critical individual lives at this location in the future, the annual lung dose is estimated to be 10.7 mrem /yr based on the existing release rates from plant operation. This dose is still well below the 25 mrem limit. The staff estimated that the release rate of airborne e#fluents would have to be increased to about 7.05 mC/yr in future operations in order to exceed the 25 mrem limit for an infant living at the nearest site boundary. To insure compliance with.the regulations, the staff proposes a license condition requiring an action level on effluent releases. Since the liquid effluent is not the major pathway in individual dose calculations, the action level will be based on measured airborne releases. Even though the staff's analysis shows that an effluent release of over 7,000 uCi/yr would be necessary to exceed the 25 mrem limit to the critical individual at the nearest site boundary, considering present release rates of approximately 3,100 uCi/yr it is the staff's opinion that in order not to violate principles of ALARA, a somewhat lower action level should be defined. The proposed action level will be a reporting requirement unless circumstances warrant more strict enforcement. The staff then recommends an action level on the release of airborne effluents to be at 1,500 uCi of U per quarter which is equivalent to an annual lung dose to an infant at the nearest site boundary of about 21.3 mrem /yr. Accordingly, the staff reconr. ends that the following conditions be added to the license: 1. If the radioactivity in plant gaseous effluents exceeds 1500 uCi per calendar quarter, the licensee shall, within 30 days, prepare and submit to the Commission a report which identifies the cause for exceeding the limit and the corre the licensee to reduce release rates.ytive actions to be taken by If the parameters important to a dose assessment change, a report shall be submitted within 30 days which describes the changes in parameters ang includes an estimate of the resultant change in dose commitment 2. In the event that the calculated dose to any member of the public in any consecutive 12-month period is about to exceed the limits specified in 40 CFR 190.10, the licensee shall take immediate steps to reduce emissions so as to comply with 40 CFR 190.10. As provided in 40 CFR 190.11, the licensee may petition the Nuclear Regulatory Commission for a variance from the requirements of 40 CFR 190.10. If a petition for a variance is anticipated, the licensee shall submit the request at least 90 days prior to exceeding the limits specified in 40 CFR 190.10.

15 References 1. Envirenmental Protection Agency, " Title 40 - Protection of the Environment, Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations," Federal Register 42(9): 2558-2561 (January 13, 1977). 2. Westinghouse Electric Corporation, Environmental Systems Department, Environmental Evaluation for Westinohouse Nuclear Fuel Division, Columbia, Soutn Carolina, Occket No. 70-1151, Pittsburgn, Pennsylvania, (March 1, 1975). 3. STAR Program for On-Site Data Diffusion Climatology, WESD, Monroeville, Pennsylvania, (1972). 4. " Meteorology and Atomic Energy," David H. Slade, Editor, USAEC, Division of Technical Information, pp. 97-104, (July 1968). 5. Snydar, W. H., and R. E. Lawson, Jr., " Determination of a Necessary Height for a Stack Close to a Building--A Wind Tunnel Study," Atmosoheric Envi ronment, Vol. 10, pp. 683-691, Pergammon Press, (1975). 6. 40 CFR 190, Environmental Radiation Protection Requirements for Normal Operations of Activities in the Uranium Fuel Cycle, Final Environmental Statement, Vol. 1, pp. 143-146, USEPA, (November 1970). 7. U.S. Nuclear Regulatory Commission - Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Office of Standards Development, (July 1977). 8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," (March 1976). 9. Task Group of Committee 2, ICRP, Task Grouo on Lung Dynamics for Committee II of the ICRP, Health Physics, Vol. 12, (1966). 10. Task Group of Committee 2, ICRP, The Metabolism of Comoounds of Plutonium and Other Actinides, ICRP Publication 19, Pergammon Press, Oxford, (1972). 11. Houston, J. R., D. L. Strengh, and E. C. Watson, DACRIN - A Computer Program for Calculating Organ Oose from Acute or Chronic Radionuclides Inhalation, BNWL - B-389, Battelle Pacific Northwest Laboratories, Richland, Washington, (1975). 12. M. H. Momeni, Y. Yuan and A. J. Zielen, The Uranium Dispersion and Dosimetry (UDAD) Code, NUREG/CR-0553, ANL/ES-72, Version IX, (1979).

13. NUREG-0172, Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake, BNWL, (November 1977).}}
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