ML19296A774
| ML19296A774 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 01/28/1980 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML19296A770 | List: |
| References | |
| NUDOCS 8002190024 | |
| Download: ML19296A774 (15) | |
Text
Docket No:
70-1113 LICENSEE:
General Electric Company FACILITY: Wilmington Fuel Fabrication Facility, Wilmington, North Carolina
SUBJECT:
RADIOLOGICAL ASSESSMENT OF INDIVIDUAL DOSE RESULTING FROM ROUTINE OPERATION - DEMONSTRATION OF COMPLIANCE WITH 40 CFR 190 I.
Background
The EPA uranium fuel cycle standard as specified in 40 CFR 190,1 limits 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 became effective on December 1, 1979, for all uranium fuel fabrication plants used for the production of LWR fuel.
The General Electric Company's (the licensee) plant is an existing uranium oxide fuel fabrication facility which is subject to the EPA standards.
Based on the most current plant operation, emission and monitoring data, the NRC staff conducted the following radiologi-cal assessment to determine if the licensee meets the EPA's standard on fuel cycle facilities.
As a result of this assessment, an action level for radioactive effluents 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.
Description or the Facility 1.
Plant Operation - General General Electric Company owns and operates a plant for the prnduction of light water reactor fuel assemblies which is located in New Hanover County about six miles north of the City of Wilmington, North Carolina.
The plant has a nominal capacity of about 1,500 metric tons of uranium per year.
Although there are five major buildings on the site, the fuel manufacturing operations (FMO) building is the only one in which uranium is processed and is thus the point at which all radioactive waste is generated.
2.
Chemical Process The raw material for fuel production is uranium enriched up to 4% U-235 in the form of uranium hexafluoride (UFs) or not more than 6% U-235 in the form of uranium dioxide (U0 ).
The U0 2
2 is used primarily for batch enrichment adjustment operations.
The UFs is converted to UO2 using either the ammonium diuranate (ADU) process or an alternate " direct conversion" pr' cess.
The major fraction of the UFs is converted to UO2 with the ADU prscess, but with
~
either process the UFs is received in 2.5 ton cylinders and introduced to the system by electric heating and vaporization.
In the ASU process, the UFs is hydrolyzed to uranyl fluoride (U0 F ) and hydrofluoric acid (HF) by piping the 22 8002190 O
2 ufo into a tank of water.
Ammonium diuranate is precipitated from the UO F 2 2 by the addition of ammonium hydroxide.
The ADU precipitate is separated from the matrix liquid by two stages of centrifugation and the aqueous effluent sent to waste disposal via quarantine tanks.
The ADU is continuously fed into a horizontal gas-fired defluorinator-calciner where it is dried at about 1,200 F and reduced to U 0s.
This compound is further reduced to UO in a 3
2 second, similar furnace.
Offgas from the defluorinator is scrubbed to remove uranium compounds and roJted to the main scrubber filter ventilation system.
UO discharged from the calciner is stored as feed for the mechanical processing 2
steps described below.
In the alternate, " direct conversion" process, the gaseous UFs and oxygen are introduced to a reaction vessel in the presence of an active flame shielded from a surrounding reducing gas.
The resulting uranium dioxide is oxidized to a higher oxide of uranium upon further conversion of the residual reducing gas to its oxidized form.
The byproduct HF is separated from the uranium oxide and collected for disposition as a salable product.
3.
Mechanical Processing U0 Powder is pressed into pellets, which are nominally one-half inch diameter 2
by one-half inch long.
After sintering in a reducing atmosphere at s1,800 C, the pellets are ground to standard diameter, purity tested, dried and loaded into Zircaloy tubes, one end of whichhas been fitted with a welded end plug.
The second end plug is then welded in place and the loaded tubes, now called rods, are assembled in 49 or 64-rod bundles (fuel assemblies) and packaged for shipment to the customers' nuclear reactor sites.
B.
Waste Confinement and Effluent Controls 1.
Gaseous Effluents Wilmington Operations involving the use of radioactive materials in unsealed physical forms are limited to low enriched uranium in the fuel manufacturing facilities, or the associated analytical laboratory.
The ventilation systems installed in these facilities are designed so that all the air from zones used to handle or process uranium is treated to remove essentially all the uranium prior to release.
Filtration is the predominant method for removing particulate uranium from discharge air streams.
High efficiency particulate air (HEPA) filters are used for this removal and are more than 99.97% efficient for 0.3 micron diameter particles.
All air exhaust systems from the uranium processing areas contain at least one stage of HEPA filters.
The four exhaust system stacks which serve the wet chemical process area have a wet scrubber installed ahead of the HEPA filters.
The air stream downstream of the scrubber is heated prior to passing through the HEPA filters.
The bleed-off from the scrubber recirculating water is routed to the uranium conversion process where any entrained uranium is recovered and reused.
The remaining process area exhaust stacks have two HEPA filters in series, and in some cases a roughing filter to reduce the loading on the primary HEPA filter.
A portion of the air from cach zone is recirculated in order to conserve heating and cooling energy and the systems are designed to provide air flow from areas of lesser contamination to areas of higher contamination.
~..
3 Each release stack is equipped with a sampling device which continuously draws a sample of the stack exhaust through a low porosity filter paper.
The filter paper is removed periodically and analyzed for uranium.
The chemical processing area samples are also analyzed fnr fluorides.
Each of the separate air-handling systems appropriately combine fans, dampers, ducts, filters, pressure signals, and pneumatic and electrical control devices to control ventilation consistent with the specific requirements of the area.
Dimensions of these areas and ventilation capacities are given in Table 1.
2.
Liquid Effluents Liquids removed by centrifugation in the conversion of UFs to UO2 are classified as fluoride waste.
Each batch is collected in the Fuel Manu-facturing Operation quarantine tanks, sampled and analyzed for its uranium concentration.
Upon verification of adequate in process uranium removal (including recycling whenever necessary), these liquids are pumped to a 65,000 gallon storage tank.
Liquids from this tank are pumped to a 100,000 gallon tank in the waste treatment facility.
This tank is constructed with a conical (funnel shaped) bottom which effects a further settling of particulate residue as the material is pumped through.
The settled solids are periodically removed and returned for rework and recovery of the uranium.
The larger, liquid fraction is subjected to further batch treatment by the addition of lime, thus precipitating the fluorides as calcium fluoride and freeing ammonia for subsequent removal by aeration.
Following an ammonia recovery operation, the resulting slurry of calcium fluoride, water, and residual ammonia is discharged to series of lagoons in which the calcium fluoride settles, residual ammonia is dissipated, and the clarified water overflows to the creek and is combined with storm water and sanitary effluents.
The State of North Carolina has approved the system which adjusts the pH of the treated process effluent to a range of 6.0 to 6.9 by the use of sulfuric acid (H 50 ).
2 4 Liquid wastes resulting from the fuel manufacturing uranium purification system (UPS) and nitric acid residues from periodic equipment cleaning operations are collected in quarantine vessels and are classed as nitrate waste.
At the waste treatment area, lime is added to precipitate the calcium uranate.
The precipitated slurry is centrifuged to remove the bulk of the calcium uranate which is returned to fuel manufacturing for recovery.
The clarified nitrate solution is routed to nitrate storage basins for concentration.
Waters from protective clothing, washing machines, laboratory sinks, floor washings, equipment decontamination, and similar fuel building services are collected in radwaste accumulator tanks.
After centrifugation to remove solids, the clarified water is sampled, analyzed, and, when released, is pumped to the general waste system where it is treated with other effluents.
4 Table 1 VENTILATION CAPACITIES OF FUEL BUILDING AREAS Volume Exhaust Recirculation No. of Air Area (cu. f')
(scfm)
(scfm)
Changes /hr UFs-UO2 Conversion 422,900 35,750 80,400 16.5 Vaporization 71,000 6,000 9,000 12.7 North Sintering 160,500 12,000 53,000 24.3 South Sintering
- 67,600 7,370 17,630 22.2 North UO Power 2
Storage 78,500 5,000 7,000 9.2 South U0 Powder 2
Storage 67,700 10,000 2,000 10.7 North Segmentizing 160,000 11,700 16,000 10.4 South Segmentizing*
74,000 4,540 7,400
- 9. 7 North Pelletizing 67,620 8,600 11,300 17.6 South Pelletizing
- 49,245 6,300 8,300 17.8 North Waste Retention 62,500 4,000 8,000 11.6 South Waste Retention 62,500 4,000 8,000 11.6
^ Expansion part of fuels building
t..
5 The recovered solids are returned to the process via the tranium purification system or stored, awaiting further recovery.
The fuel manufacturing operation generates about 20,000 gallons per day of nitrate-bearing liquide averaging less than 2 ppm uranium at an average enrichment of 2.2 percen U-235.
These liquids are transported by tank truck to the plant of the Federal Paper Board Company at Riegelwood, North Carolina, about 20 miles from the GE plant.
At Riegelwood, the weak ammonium nitrate solution is fed into the waste treatment system where it helps to provide needed nitrogen for promotion of bacterial decomposition of the paper board waste.
The uranium in the GE waste stream is diluted by the approximately 40 million gallons of water per day used at the Federal Riegelwood plant so that the calculated final uranium concentration at the outfall to the Cape Fear River from the Riegelwood waste treatment system is about 0.5 ppb.
3.
Solid Wastes Solids contaminated with radioactive materials are stored on site awaiting treatment and/or shipment.
Uranium-contaminated materials are stored in NRC-approved containers within the exclusion area.
Uranium-contaminated solid wastes, which contain amounts of uranium larger than desirable to discard, are held for uranium recovery.
Solid wastes held for reprocessing are sealed and labeled.
Radioactive solid wastes are disposed of by a private waste disposal contractor who is licensed and equipped to manage such wastes.
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 2 and 3 summarize the results on the radioactivity measured in air and liquid effluent discharged to the environment for the past few years (1976-1978).
D.
Description of the Site Environment Related to Radiological Assessment at the Maximum Nearest Resident 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.
1.
Site Location The Wilmington uranium fuel fabrication facility is located in New Hanover County, 6 miles north of the City of Wilmington.
This location is in the southwestern corner of the coastal plains region of North Carolina.
2.
Land and Water Uses The site contains 1,664 acres of land, none of which is an irretrievable commitment.
The land around the site is heavily timbered.
The population
6 Table 2 SEMIANNUAL AIRBORNE RELEASES (Ci)
Period U-234 U-235 U-236 U-238 Total U July 1, 1976-June 30, 1977 2.52x10[3 July 1, 1977-Dec. 31, 1977 1.33x10.5 6.57x10_4 6.70x10.4 Jan. 1, 1978-June 30, 1978 4.19x10 4 2.79x10 5 6.09x10 6 1.81x10 4 6.34x10 4
~
July 1, 1978-Dec. 31, 1978 1.48x10 3 1.02x10 4 2.25x10 5 7.39x10 4 2.34x10 3
~
6 Month Average 1.23x10 3 Twice 6 Month Averagc 2.47x10 3 Table 3 SEMIANNUAL LIQUID RELEASES (Ci)
Period U-234 U-235 U-236 U-238 Total U July 1, 1976-June 30, 1977 7.6x10 1 July 1, 1977-Dec. 31, 1977 l.48x100 Jan. 1, 1978-June 30, 1978 9.90x10.2 6.60x10.3 1.44x10.3 4.29x10_2 1.50x10_1 July 1, 1978-Oec. 31, 1978 1.46x10 1 1.00x10 2 2.22x10 3 7.28x10 2 2.31x10 1 6 Month Average 5.24x10 1 Twice 6 Month Average 1.05x100
7 distribution within a five-mile radius is shown in Figure 1.
As shown in the figure, the area surrounding the site is sparsely populated with farms, single family dwellings and limited commercial activities located primarily along major highways.
Liquid effluent is released from the site via a series of waste lagoons to the site creek after sampling and analysis.
The receiving water is the Northeast Cape Fear River at the western boundary of the plant site.
All site surface storm water and treated waste drainage flows into the river.
The flow character-istics of the Northeast Cape Fear River, near the plant site, are complex because of its estuarine characteristics with dilution and transit times a function of both tidal influences and fresh water inflow.
3.
Diffusion Climatology Onsite meteorological data on wind speeds and direction is not available from the licensee.
However, general climatological characteristics in the area can be referenced to the U.S. Weather Bureau recording station at Wilmington, North Carolina which is about six miles south of the site.
For the atmospheric dispersion calculations, joint frequency distributions of wind speed by stability class were calculated using the STAR 2 program based on observations made at Wilmington.
The meteorological dispersion factors (X/Q),
were produced from the Gaussian Plume model and diffusion coefficients for Pasquill type turbulence using a computer code generated as in Regulatory Guide 1.111.a.s In evaluating the annual average X/Q values, a ground level release was conservatively assumed with correction for building wake effects.4 The annual average X/Q's as a function of distance up to 50 miles from the site in the sixteen 22-1/2 degree compass point sectors (i.e., centered on the north, northeast, southeast, etc.) were calculated and are shown in Table 4.
4.
The Nearest Resident Figure 1 shows the plant site and population distribution within 5 miles.
The nearest resident, for which the evaluation was conducted, is located approxi-mately 600 meters south-southeast of the plant site.
The annual average X/Q values shown in Table 4 indicate that at the same distance, an individual (if any exists at that location) in some other direction may be subject to a higher dose due to less diffusion of the effluent.
If, in the future, the nearest residence changes to an area of greater potential dose than the present, the analysis will be redone for that location.
E.
Environmental Impact from Routine Plant Operation 1.
Methodology for Radiological Assessment The general approach to demonstrate compliance with the dose limits of the standard is as follows:
(i) Effluents released from plant operation will be monitored to determine the quality of radionuclides discharged into the environment.
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10 (ii) Environmental dose moaels developed by the NRC will be used to estimate dose commitment rates from all significant pathways.
Because of the closeness of the nearest site boundary (500 ft), it is the staff's recommendation that a detailed environmental monitoring program be required to supplement the routine 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 Requiremerits for Normal Operations of Activities in the Uranium Fuel Cycle.5 The source terms (radioactive effluent release rates) from the Wilmington plant are measured values.
The atmospheric dispersion model is based on Regulatory Guide 1.111.6 Other environmental pathways and models are based on Regulatory Guide 1.1097 with the exception that for the inhalation pathway, dose conversion factors for various organs were generated using the ICRP Task
~11 Group Lung Model.8 The dose conversion factors from the Task Group Lung Model depend on the particle size and solubility of released radioactive compounds.
If this information is not available from the licensee, a reasonable, conservative approach will be applied for the radiological impact assessment.
For example, the particle size is assumed to have an average diameter (AMAD) of 0.3 micrometers for particles passing through HEPA filters and 1.0 micrometers AMAD for particles not passing through HEPA filters.
The released particles are further assumed first to be completely in an insoluble form which will provide a calculated maximum lung dose for the inhalation pathway and then completely in a soluble form which will provide a calculated maximum bone dose for the ingestion pathway.
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 effluent characteristics data for a more realistic evaluation.
2.
Maximum Individual Dose The radiological impacts are assessed by calculating the maximum individual dose to the closest resident, who is living at about 600 meters to the south-southeast.
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 chronic intake of radionuclides which will accrue during the remaining lifetime (50 years) of the individual.
For airborne effluents released into the environment, the pathways considered in the individual dose estimates include (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 include (a) aquatic food (fish), and (b) shoreline deposition.
The models and various assumptions involved in the above environmental pathways can be referred to in greater detail in Regulatory Guide 1.109.
Table 5 summarizes the results of the estimated maximum annual dose from airborne and liquid effluents to the nearest resident.
11 As shown in Table 5, the critical pathway is due to inhalation resulting in a maximum dose to the lung of 1.89 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., 3.4 mrem /yr, less than 14% of the environmental standard. 23 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 will be no adverse effect due to the release of effluents from normal plant operation.
Table 2 summarizes the semiannual release rates of radiological airborne effluents which were used as source terms for this assessment.
The release rates are measured values.
The respective semiannual release rates were averaged for a representative six-month rate and that value doubled for the annual release value shown in the tables and used in the calculations.
For liquid effluents discharged into the Northeast Cape Fear River and then to the Cape Fear River, it was conservatively assumed that the uranium is in a soluble form.
It was further assumed that the liquid release was only diluted by the river flow at the point of release to tne Northeast Cape Fear River.
F.
Conclusion and Recommendation The normal operation of the GE fuel fabrication plant results in the release of a minute quantity of radioactivity into the environment.
Based on past operation, the annual release of racioactivity included approximately 2.47 millicuries of uranium in airborne effluents and 1.05 curies of uranium in liquid effluents.
The nearest resident is located at about 600 meters south-southwest of the plant site.
The annual lung dose to the critical individual at the nearest resident was estimated under conservative assumptions to be 3.4 mrem /yr which represents less than 14% of the 25 mrem limits 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 routina operation of the Hematite fuel fabrication plant.
The staff recognizes that the nearest resident located 600 meters south-southwest of the facility might not represent the potential maximum impact from the Wilmington operation.
The staff estimated that the maximum impact in the unrestricted area could be at the nearest site boundary, 500 feet to the_ south of the center of the Fuel Building. The X/Q at this location is 6.92x10 5 3
sec/m.
If a critical individual were to live at the nearest site boundary in the future, conservative calculations predict that an annual dose of greater than 39 mrem could be imposed.
Accordingly, in order to evaluate the real dose to an individual at the fenceline, an environmental monitoring program will be required for a period of one year to determine the concentration, particle size, and solubility of any radionuclides at the site boundary.
Upon completion of the evaluation of the real dose for an individual at the fenceline using the above measured values, the staff will impose continuing appropriate environmental monitoring requirements as conditions of your license.
.2 TABLE 5 ESTIMATED MAXIMUM ANNUAL DOSE FROM AIRBORNE AND LIQUID EFFLUENTS TO THE NEAREST RESIDENT Pathways Organ Dose (millirem /yr)
A.
Air Effluents Total-Bocf Lung Bone
~
1.
Direct Irradiation 2.38x10_4 2.
Direct Inhalation
- 9.86x10 4 1.89 1.59x10 2 3.
Ingestion Due to Airborne Deposition a.
Vegetables **
4.57x10:
7.38x10 3
2 b.
Meat 1.86x10.5 4
c.
Milk 7.71x10 5 3.01x10_a 1.25x10 B.
Liquid Effluents 1.
Potable Water No Pathway 2.
Aquatic Food (Fish) 3.58x10_a 5.79x10 2 3.
Shoreline Deposition 9.09x10 7 Total (millirem /yr) 9.47x10 3 1.89 1.49x10 1
^ Assume 80% residence time.
- Includes nonleafy and leafy vegetables.
Since site specific information is not available, the staff assumed 76% of the produce and 100% of leafy vegetables consumed are grown in the garden of interest.
~
13 The regulation is directed to the protection of the nearest resident from routine releases and cction levels will be based on the effect of effluent releases to the nearest real person. Since the location of the nearest resident is critical to the facility's ability to remain in compliance with the regulations (less than 25 mrem /yr), it is necessary and will be a con-dition of the license that the NRC be notified of any change in location of the nearest resident. Similarly, notification of any other encroachment of the site boundary will be required.
Therefore, to ensure compliance with the regulations, the staff proposes the following license condition requiring an action level on effluent releases.
Even though the staff's analysis shows that an effluent release of over 18.1 mci /yr would be necessary to exceed the 25 mrem limit to the critical individual at the nearest residence, considering present release rates of approximately 2.5 mci /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 1250 pCi of U per quarter which is equivalent to an annual lung dose to an infant at the nearest residence of about 7 mrem /yr. Accordingly, the staff recommends that the following conditions be added to the license:
1.
If the radioactivity in plant gaseous effluents exceeds 1250 pCi 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 corrective actions to be taken by the licensee to reduce release rates.8 If the parameters important to a dose assessment change, a report shall be submitted within 30 days which describes the changes in parameters and cludes 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 exccedng the limits specified in 40 CFR 190.10.
3.
Air sampling stations shall be installed and operated in the prevailing wind directions, i.e., N, S, SW and NE (location should be near the fenceline if possible). Air samples shall be collected continuously and in addition to the gross-alpha analysis, the samples at each location shall be ccmposited and analyzed on a monthly basis for uranium isotopes with an analytical sensitivity of at least 10-16 pCi/ml.
I The report or petition should be submitted to the Director, Office of Nuclear Material Safety and Safeguards with a copy to the Director of the Regional Office of the Office of Inspection and Enforcement.
14 4.
If the results of the sampling program indicate an airborne urinium concentration of greater than 3.45 x 10-15 3
Ci/M, the licensee shall, within 30 days, institute an air sampling -
program at the site boundary in the direction of the nearest resident to determine the particle size distribution of radioactive material with a cascade impactor of multiple stages covering non-respirable and respirable particle size ranges.
The particle size distribution analysis may be performed on a semi-annual basis.
5.
Samples taken at the station in the southern location, shall be composited quarterly and analyzed for uranium solubility.
The solubility analysis shall follow the methodology and 2
procedures established by BNWL,3 or an equivalent method acceptable to the NRC.
If a laboratory other than BNWL is used for the analysis, the licensee shall provide the NRC with a split sample so that the NRC can perform a verification analysis.
2Solubility Classification of Airborne Products from Uranium Ores and Tailings Piles - D. R. Kalkwarf, BNWL, November 1978.
3Second Quarterly Report on Solubility Classification of Airborne Products from LWR-Fuel Plants - D. R. Kalkwarf, BNWL, October 15, 1979.
15 REFERENCES 1.
Environmental Protection Agency, " Title 40 - Protection of the Environment, Part 190 - Environmental Radiation Protection Standards for Nuclear Power p
Operations," Federal Register 42(9):
2858-2861 (January 13, 1977).
2.
STAR Program for On-Site Data Diffusion Climatology, WESD, Monroeville, Pennsylvania, (1972).
3.
" Meteorology and Atomic Energy," David H. Slade, Editor, USAEC, Division of Technical Information, pp.97-104, (July 1968).
4.
Snyder, W. H., and R. E. Lawson, Jr., " Determination of a Necessary Height for a Stack Close to a Building--A Wind Tunnel Study," Atmospheric Environment, Vol. 10, pp. 683-691, Pergammon Press, (1975).
5.
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 1976).
6.
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 Raactors," Office of Standards Development, (July 1977).
7.
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).
8.
Task Group of Committee 2, ICRP, Task Group on Lung Dynamics for Committee II of the ICRP, Health Physics, Vol. 12, (1966).
9.
Task Group of Committee 2, ICRP, The Metabolism of Compounds of Plutonium and Other Actinides, ICRP Publication 19, Pergammon Press, Oxford, (1972).
10.
Houston, J. R., D. L. Strengh, and E. C. Watson, DACRIN - A Computer Program for Calculating Organ Dose from Acute or Chronic Radionuclide Inhalation, BNWL - B-389, Battelle Pacific Northwest Laboratories, Richland, Washington, (1975).
11.
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).
12.
NUREG-0172, Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake, BNWL, (November 1977).