ML20053A817
| ML20053A817 | |
| Person / Time | |
|---|---|
| Site: | 07001100 |
| Issue date: | 05/31/1982 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20053A808 | List: |
| References | |
| NUDOCS 8205270312 | |
| Download: ML20053A817 (78) | |
Text
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ENVIRONMENTAL IMPACT APPRAISAL OF THE PROPOSED SPECIAL NUCLEAR MATERIAL LICENSE RENEWAL related to THE NUCLEAR FUEL FABRICATION PLANT COMBUSTION ENGINEERING, INC.
WINDSOR, CONNECTICUT Docket No. 70-1100 May 1982 U.S. Nuclear Regulatory Comission Office of Nuclear Material Safety and Safeguards Division of Fuel Cycle and Material Safety Washington, D.C.
8 2 0 5 2 7p3/ a
CONTENTS Page LIST OF FIGURES y
LIST OF TABLES..........................
vii 1.
PURPOSE OF AND NEED FOR ACTION 1-1
1.1 INTRODUCTION
. l.
1-1 1.2
SUMMARY
OF THE PROPOSED ACTION.............
1-1 1.3 NEED FOR ACTION 1-2 1.4 THE SCOPING PROCESS 1-2 2.
ALTERNATIVES INCLUDING THE PROPOSED ACTION 2-1 2.1 THE ALTERNATIVE OF NO LICENSE RENEWAL 2-1 2.2 THE ALTERNATIVE OF LICENSE RENEWAL WITHOUT CHANGE 2-1 2.3 EXTERNAL APPEARANCES..................
2-1 2.4
SUMMARY
OF OPERATIONS 2-3 2.5 PRESENT OPERATION 2-5 2.5.1 Pellet fabrication 2-5 2.5.2 Fuel rod assemblies...............
2-6 2.5.3 Quality assurance operations 2-6 2.5.4 Fuel bundle assembly 2-6 2.5.5 Scrap recovery 2-7 2.5.6 Shipping 2-7 2.5.7 Nuclear laboratories facility, Buildino 5....
2-7 2.5.8 Radiological waste treatment......... '.
2-9 2.5.9 Industrial wastewater treatment.
2-10 2.5.10 Sanitary wastewater treatment..........
2-11 2.6 WASTE CONFINEMENT AND EFFLUENT CONTROL.........
2-12 2.6.1 Gaseous effluent 2-13
^
2.6.2 Liquid effluent..
2-15 2.6.3 Solid waste effluent 2-17 2.7 DECOMMISSIONING 2-19 2.8 MATERIAL CONTROL AND ACCOUNTABILITY 2-19 2.9 STAFF EVALUATION OF THE PROPOSED ACTION AND ALTERNATIVES......................
2-20 REFERENCES FOR SECTION 2 2-21 3.
AFFECTED ENVIRONMENT 3-1 3.1 SITE DESCRIPTION....................
3-1 3.2 CLIMATOLOGY AND METEOROLOGY 3-1 3.2.1 Winds, including atmospneric discersion.....
3-1 3.2.2 Precipitation..................
3-2 3.2.3 Severe weather 3-3 3.3 AIR QUALITY 3-3 3.4 SOCI0 ECONOMICS.....
31 3.5 LAND USE.......
32 iii
Page 3.6 WATER 3-6 3.6.1 Surface wa te r..................
3-6 3.6.2 Groundwater...................
3-7 3.7 GEOLOGY, MINERAL RESOURCES, AND SEISMICITY.......
3-9 3.7.1 Geology.....................
3-9 3.7.2 Mineral resources................
3-9 3.7.3 Seismicity 3-10 3.8 ECOLOGY 3-10 3.8.1 Terrestrial...................
3 3.8.2 Aquatic.....................
3-14 3.9 RADIOLOGICAL CHARACTERISTICS (BACKGROUND) 3-15 REFERENCES FOR SECTION 3 3-16 4
ENVIRONMENTAL CONSE0VENCES OF PROPOSED LICENSE RENEWAL 4-1 4.1 MONITORING PROGRAMS AND MITIGATORY MEASURES 4-1 4.1.1 Onsite monitoring program............
4-1 4.1.2 Offsite monitoring program 4-5 4.1.3 Mitigating measures..............,
4-7 4.2 DIRECT EFFECTS AND (HEIR SIGNIFICANCE 4-7 4.2.1 Air quality...................
4-7 4.2.2 Land use 4-8 4.2.3 Water......................
4-9 4.2.4 Ecological 4-9 4.2.5 Radiological impacts 4-11 4.3 INDIRECT EFFECTS AND THEIR SIGNIFICANCE 4-16 4.3.1 Socioeconomic effects..............
4-16 4.3.2 Potential effects of accidents 4-17 4.3.3 Possible conflicts between the proposed action and the objectives of federal, regional, state, and local plans and policies 4-23 4.3.4 Effects on urban quality, historical and cultural resources, and society.........
4-23 REFERENCES FOR SECTION 4 4-24 Appendix A.
METHODOLOGY AND ASSUMPTIONS FOR CALCULATING RADIATION DOSE COMMITMENTS FROM THE RELEASE OF RADIONUCLIDES A-1 iv
LIST OF FIGURES Figure Page 2.1 Plot plan of the CE Windsor site 2-2 2.2 Nuclear fuel cycle 2-4 2.3 Nuclear fuel fabrication facility, Building 17, general layout floor plan.................
2-5 2.4 Typical block diagram for fuel bundle assembly for the nuclear fuel facility plant, Building 17 2-8 2.5 Sanitary waste treatment flow sheet, Building 10, serving Buildings 5 and i7~
2-11 3.1 CE Windsor site......................
3-2 3.2 Water balance diagram for Buildings 5 and 17 on an annual basis 3-8 3.3 A preliminary map of horizontal acceleration in rock with 90% probability of not being exceeded in 50 years........................
3-11 3.4 Diagrammatic cross section of Northwest Park showing the typical toposequence of plant communities in the CE Windsor site area 3-13 4.1 Onsite sampling locations.................
4-2 4.2 Offsite sampling locations 4-4 A.1 Pathways for exposure of man from releases of radioacti ve ef fl uents................... A-4 i
y
LIST OF TABLES Table Page 2.1 Summary of Combustion Engineering industrial effluent data for 1980 2-10 2.2 Average isotopic mixture 2-13 2.3 Estimated airborne uranium releases............
2-14 2.4 Discharge limits - process gases 2-14.
2.5 Chemicals used at the fuel fabrication plant 2-18 3.1 Summary of air quality data for the Combustion Engineering Windsor site 3-4 3.2 Approximate percentage land use in the townships of Windsor, Bloomfield, and East Granby..........
3-5 3.3 Selected water quality parameters of the Farmington y
River in the vicinity of :he Combustion Engineering Windsor plant site 3-7 3.4 Vegetation types of Northwest Park near the Combustion Engineering Windsor site............
3-12 3.5 Fish species occurring in the Farmington River in the vicinity of the Combustion Engineering Windsor site 3-14 4.1 Onsite environmental monitoring program for the CE Windsor plant 4-1 4.2 Summary of nonradiological effluents from surface l
water sampling stations shown in Fig. A.1.........
4-3 4.3 Offsite environmental monitoring program for the Combustion Engineering Windsor plant
...........'4-6 4.4 Nonradiological emissions to tne atmosphere........
4-8 4.5 Summary of liquid chemical effluents 4-10 4.6 Summary of Combustion Engineering industrial drain discharge data for NPDES permit requirements 4-11 4.7 Annual release rate of radienuclides in the stack effluents of the Combustien Engineering Windsor fuel fabrication facility.................
a-12 vii
Table Page 4.8 Fifty-year dose commitment to the maximally exposed individual at the nearest residence from the airborne effluents of the Combustion Engineering Windsor plant...........................
4-13 4.9 Fifty-year dose comitment to the maximally exposed individual at the nearest boundary from the airborne effluents of the Combustion Engineering Windsor plant...........................
4-14' 4.10 Fifty-year dose commitment from airborne effluents to the population living within 80 km of the Combustion Engineering Windsor plant 4-14 4.11 Annual release rate of radionuclides in the liquid effluents from the Combustion Engineering Fuel Fabrication Facility and the concentration in the Farmington. River 4-15 4.12 Maximum 50-year dose commitment to individuals from the routine release of liquid effluents into the Farmington River.................
4-16 4.13 Calculated maximum dose potential at the nearest residence from airborne radionuclides resulting from a criticality event of 1019 fissions.........
4-17 A.1 Dose conversion factors for external exposure pathways A-5 A.2 Dose conversion factors for internal expost.re pathways A-6 A.3 Intake parameters (adult) used in lieu of site-specific data A-7 l
A.4 Ground-level CHI /Q values at various distances
(
in each compass direction.
A-9 A.5 Frequency of atmospheric stability classes for each compass direction A-10 A.6 Frequencies of wind directions and true-average wind speeds........................
A-ll A.7 Other parameters used in determining exposure to air concentrations of radionuclides released in the building vent effluents..................
A-li l
A.3 Peculation distribution by distance and sector used to calculate doses given in Table.1.10............
A-12 l
viii
1.
PURPOSE OF AND NEED FOR ACTION
1.1 INTRODUCTION
The Combustion Engineering, Inc. (CE), nuclear fuel fabrication facility in Windsor, Connecticut, produces fuel assemblies for use in comercial light-water reactors.
The facility processes low-enriched uranium
(<4.1% 235U) in the form of uranium dioxide (U0 ) powder into sintered 2
pellets, encapsulates them in metal tubing, and clusters the tubing into fuel assemblies for shipment.
This environmental appraisal is in response to an application by CE for renewal of Special Nuclear Material (SNM) License No. SNM-1067 covering plant operations at Windsor, Connecticut.
It is prepared by the U.S.
Nuclear Regulatory Comission (NRC))with the technical assistance of the Oak Ridge National Laboratory (0RNL pursuant to the Council on Environ-mental Quality (CEQ) regulations (40 CFR Parts 1500-1508) and the NRC regulation (10 CFR Part 51), which implement requirements of the National Environmental Policy Act (NEPA) of 1969 (P.L.91-190).
Paragraph 1508.9 of the CEQ regulations (40 CFR) defines " environmental assessment" as follows:
1.
An environmental assessment is a concise public document for which a federal agency is responsible and which serves to briefly provide sufficient evidence and analysis for e
determining whether to prepare an Environmental Impact Statement (EIS) or a finding of no significant impact, aid an agency's compliance with the Act when no EIS is e
necessary, and facilitate preparation of a Statement when one is necessary.
2.
An environmental assessment shall include brief discussions of the need for the proposal, of alternatives as required by Sect.102(2)(E),
of the environmental impacts of the proposed action and alternatives, and a listing of agencies and persons consulted.
An Envircnmen:ci *rycc: D w n :l che Carh a:icn =ngineering, :".c.,
Ac ecr Ekel Fabricaric-F: :.; icy was issued in October 197a.
There-fore, only material 9,,re '. to the proposed apolication for renewal of SNM-1067 is repu; ; )
lic environmental appraisal.
1.2
SUMMARY
OF THE PROPOSED ACTION The proposed action for whicn this environmental acoraisal is cerformed is the routine renewal of the lice 3se necessary for CE to continue l
oceration.
Princical activities in the 'abription facility include the l
processing of low-enricned uranium (<a.15 235U) in tne form of UO: Ocwder l
l i-1
1-2 into pellets, encapsulating the pellets in metal tubing, and clustering the tubing into fuel assemblies for shipment to nuclear reactor sites.
In addition to the fabrication facility, there is an onsite supporting nuclear laboratory, radiological treatment area, and a sanitary waste-water treatment facility.
The CE site comprises approximately 225 ha (556 acres), of which about 7.1 ha (17.5 acres) is occupied by the fabrication plant and supporting activities.
The site is located about 14.5 km (9 miles) north of Hartford, Connecticut.
1.3 NEE FOR ACTION Among the ilternative actions available to NRC is the denial of renewal of the SNh license of the applicant.
However, the denial of renewal of the SNM license would be considered only if issues of public health and safety cannot be resolved to the satisfaction of the re';ulatory authorities involved.
1.4 THE SCOPING PROCESS The overall operations and impacts of the CE facility were appraised in October 1974.
Because of existing documentation and the very limited impacts associated with the operation of this facility, the NRC staff determined that a formal scoping process was unnecessary to adequately assess the proposed action.
In conducting this appraisal, the staff met with the applicant to discuss items of information related to facility operations and to seek additional information that might be needed for an adequate appraisal.
In addition, the staff sought information from other sources to assist in the evaluation and conducted field inspections of the project site.
2.
ALTERNATIVES INCLUDING THE PROPOSED ACTION 2.1 THE ALTERNATIVE OF NO LICENSE RENEWAL Not granting a license renewal would cause CE to cease those fuel fabricating processes described previously. This alternative would be considered only if issues of public health and safety could not be resolved.
The only benefits to be gained by such a course of action would be the cessation of the environmental impacts (as described in Sect. 4) that have been determined to be acceptably small.
2.2 THE ALTERNATIVE OF LICENSE RENEWAL WITHOUT CHANGE This alternative, which is the proposed action, would result in the continued operation of the CE facility essentially as it has been operated for the past 11 years.
2.3 EXTERNAL APPEARANCES Of the 225 ha (556 acres) owned by CE, only a total of 7.1 ha (17.5 acres) of the land is occupied by buildings and support facilities.
The fuel fabrication facility and nuclear laboratories, licensed by NRC to handle the special nuclear material covered by this environmental impact appraisal, occupy a total area of 1 ha (2.5 acres).
Site facilities consist of the nuclear fuel fabrication facility, nuclear laboratories, radiological waste treatment area, and sanitary wastewater treatment facility.
The nuclear fuel fabrication facility, Building 17, uses 3900 m2(42,000 2
ft ) of floor space in the manufacturing and admin-istration of nuclear fuel assemblies.
The facility is 37 m (120 ft) wide by 104 m (340 ft) long by 9.1 m (30 ft) high on a concrete floor; the facility has corrugated asbestos siding and poured-gypsum roof deck.
The administration area is partial brick constrtction.
The nuclear laboratories, Building 5, use 5574 m2 2
(60,000 ft ) of floor space in a multiple-level structure with a main bay and three wings (north, central, and south).
Attached to the central wing of the building is a hign-bay test facility 30 m (100 ft) high.
The main bay and the three wings of Suilding 5 contain office space that occupies a total of 2508 m2 2
(27,000 ft ).
Mechanical testing and research and development (R&D) work occupies 1533 m2 2
(16,500 ft ) in the main bay; central wing, including nigh-bay area; and south wing.
The electronics testing, fabrication of special research test fuel, and chemistry testing labcratory of croduc-tion nuclear fuel for quality ;ontrol purcoses occupy the remaining areas of tne main bay and three wings.
Figure 2.1 shows locations and identifies buildings and facilities of the affected area occupied on the CE Windsor site.
2-1
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Plot plan of the CE Windsor site.
Source:
Combustion Engineering Envivorunental tuq.aa infinenation, Combustion Engineering, Inc., Windsor, Conn., USNRC Document No. 70-1100, April 1981.
2-3 The CE power system group located at the Windsor site employs approximately 5000 persons; approximately 200 of these employees work directly in the nuclear fuel fabrication area of Building 17.
2.4
SUMMARY
OF OPERATIONS The fuel fabrication facility and nuclear laboratories at CE Windsor were designed and constructed to fabricate low-enriched uranium (<4.1%
235U) fuel assemblies for use in comercial power light-water reactors.
The major operations include (1) processing the UO2 powder into pellets,-
(2) encapsulation of the pellets into fuel rods, (3) fabrication of fuel rods into final assemblies, and (4) shipment of final assemblies to the CE customer's reactor site.1 Construction of the fuel fabrication facility was begun in January 1967 and completed in December 1967.
Commercial operations began under SNM License No. 551, which limited the facility to 1000 kg of 235U when it started operating fuel rod loading in 1968.
A second SNM license was approved under authorization granted by the then-existing U.S. Atomic Energy Commission in License No. SNM-1067 in May 1968 for full-scale commercial operation (G. Johnstone, CE, personal communication to John Williams, July 24,1981).
Figure 2.2 schematically describes the general sequence of operations (nuclear fuel cycle) involved, from mining uranium are to the production of electrical energy at the power plant, including the uranium and plutonium recycle.
The boxed-in area depicts the role of the nuclear fuel manufacturing facility at Windsor (NFM-W) in this cycle.
The NFM-W is operated under the regulations of the NRC, Environmental Protection Agency (EPA), state of Connecticut's Department of Environ-mental Protection (DEP), and CE policies and procedures.
The techniques and processes incorporated into the NFM-W have been successfully demon-strated during 11 years of operation.
CE operates a nuclear fuel manufacturing facility in Hematite, Missouri (NFM-H), that converts uranium nexafluoride (UF-) to uranium dioxide (UO ) powder suitable for pressing into fuel pellets.
The facility has the capability to convert the U0- powder into finished fuel pellets suitable for encaosulating in fuel rods.
Shipments of virgin U0 2 powder and pellets are received from the NFM-H plant by the applicant in Windsor Connecticut. The Windsor facili_ty does not have the t
capability to convert UF; to U0.
The NFM-N site has an average 2
annual shipment of 200,000 kg of UO: in the form of encapsulated UO:
pellets in fuel assemolies.
CE processes various enrichments of 2 HU up to '.1% for fuel assemolies and thus recuires a comclete cleanu:: ce: ween camcaigns in :ne production facility.
2-4 ES 5902R MINING I
MILLING I
FLUORINATION I
ENRICHMENT l
CONVERSION TO URANIUM DIOXIDE ROLE OF THE
- COM8USTION ENGINEERING NFM.H H EM ATITE. MISSOU RI l
F AB RICATION
- i l
l ROLE OF THE l
- COMBUSTION ENGINEERING NFM-W I
ASSEMB LY c'
1 REACTOR CORE NUCLEAR POWER PLANT WASTES SPENT :LEL I
r lf lI l
ciSPOSAL sTO R AGE l
Fig. 2.2.
Nuclear fuel Cycle.
Source:
Adapted from.10 CFR Part 190, Fig. 3.3.
l 1
2-5 A general floor plan for the nuclear fuel fabrication facility and location of the various operations performed therein are shown in Fig. 2.3.
IS4901 I
N POWOE R STORAGE Lu CH PELLET STA
- ING, F A8 a; CATION p
UO RECYCLE LOCxERS 2
AREA FUEL SUNOLE FUEL ROO OFFICES V E N*S ASSEMBLY AREA AREA ROOM CENTRAL 04 INSPECTION TOOL CRIS l
CLEANING AND PICKLING AREA
\\
CONTAINER M ACHINE PREPARATION AREA SHOP AUTO.
WELDING LUNCH ROOM CLAVE AREA l
Fig. 2.3.
Nuclear fuel fabrication facility, Building 17, general layout floor plan.
Source:
G. Johnstone, CE, personal communication to John Williams, July 24, 1981.
Section 2.5 describes the current operations performed within the fuel fabrication facility (Building 17), nuclear laboratories facility (Building 5), and the generation of waste effluents associated with tnose production and support operations.
Section 2.5 summarizes the nature of the effluent controls and waste effluent treatment from the fuel fabrication facility and the nuclear laboratories.
Plant equip-ment is designed and operated to keep hazardous chemicals and radio-active discharges to the environment well within acplicable federal regulations.
2.5 PRESENT OPERATION 2.5.1 Pellet fabrication The U0, cowder and cellets from the chemical conversion olant in demati;e, Misscuri, are received in :ne virgin ocwder storage area of Builcing 17.
The first operation, in the case of UO: :cwcer after ta:ch weianing, is dis-charge to a blender. (The CE-Hematite cellets are storea until neeced.) In
2-6 addition, binder and lubricating chemicals are added to the blender.
The blended powder is then discharged into a granulator to produce a free-flowing powder for subsequent use in the pellet-pressing operation.
The free-flowing UO2 powder is pressed with a 27-t (30-ton) hydraulic press to the desired " green" density and geometrical shape.
It is then loaded into molybdenum-sintering furnace boats.
The " green" pellets are processed through a cracked ammonia and nitrogen gas atmosphere in a dewaxing furnace at approximately 1200 C to remove the lubricating material which was introduced into the UO2 powder to irprove the press-ing operations.
The dewaxed pellets (green) are transferred to the sintering area where they are sintered to approximately 95% of the theoretical UO2 density at a temperature of about 1800 C.
Pellets discharged from the sintering furnace are ground in a wet, centerless grinder and inspected for dimensions and density.
Rejected pellets are placed in containers and stored for subsequent recovery in the U02 recycle area.
All in-spec UO2 pellets are transferred to the bulk drying furnace operated at approximately 500 C to remove moisture pickup from the grinding operation.
Pellets are stored until needed for rod loading.
2.5.2 Fuel rod assemblies As needed, pellets are transferred from storage to the loading station where they are prestacked by aligning in a linear array.
Each pellet stack's length is sufficient to fill one fuel rod.
The stack, weigned for accountability purposes, is loaded into a fuel zirconium tube with one end cap already welded in place.
The final cap weld is made and the fuel rod is inspected to ensure integrity of encapsulation and quality of product. All rod caps are acid-pickled and cleaned before the final welding to the rods.
2.5.3 Quality assurance coerations Finished fuel rods are subjected to a variety of inspection operations to ensure the quality of the final product.
These operations include dimensional inspection, radiography, gamma scanning, and leak testing.
Following these operations, finished fuel rods are transferred to hori-ontal storage racks to await further processing in the final assembly area.-
2.5.4 Fuel bundle assemoly Comoleted fuel rods and ccmoleted fuel bundle hardware, consisting of end fittings and the fuel rod holding cage, are brought together in the assembly room.
The fuel rods are inserted into the rod holding cage, and the end fitting is attached to the end of the fuel buncle assembly.
2-7 2.5.5 Scrap recovery Scrap recovery (recycle) is characterized as recovery of either " green" or "hard" (i.e., sintered) UO2 pellets.
The block diagram in Fig. 2.4 shows where typical input to scrap recovery occurs and from what process step it was rejected.
The " green" recycle process includes dewaxing, hamermilling, cross-blending, packaging, and storing the resulting product.
The "hard" pellets must be recycled through the oxidation-reduction furnace, Micronizer,* cross-blended, and packaged for storage.
Both UO2 scrap recycle materials are reintroduced into the main process stream at the batch-weighing operation.
This recycled product is sampled and analyzed (as are any other incoming powder supplies) before release to manufacturing.
Only a small fraction (<5%) of the U02 pellets require recycling.
There are no wet chemistry steps involved in UO2 scrap recovery (G. Johnstone, CE, personal comunication to John Williams, July 24,1981).
2.5.6 Shipping Completed fuel assemblies are loaded and shipped in government-approved containers for delivery to customers' reactor sites.
Solid waste treat-ment and disposal are described in detail in Sect. 2.6.3.
There will be no liquid or gaseous effluents from the shipping operations.
Before any waste is sent out for disposal, it is measured for uranium content.
2.5.7 Nuclear labor d es facility, Building 5 o
2.5.7.1 Ceramic laboratory The ceramic laboratory is involved with the development and testing of research fuel materials and routine pellet testing for the fuel fabri-cation plant.
Various outside vendors supply CE with fuel material that is processed under different test conditions for research.
The fuel is compacted and sintered in a manner similar to that for production fuel.
The completed fuels are then tested, loaded in rods, and shipped to independent test reactors for evaluation.
Micronizer is a name in centrifugal classification equipment that uses the principle of controlling centrifugal force and air currents to grind carticles against each other on imoact to the desired fineness.
2-8 ES-5900 MAINLINE PROCESS UO POWDE R 2
RECEIVING l
FA 4 FA 4 "GRE EN" SCR A.P WEIGHING "H AR0" SCR AP BATCH RECOVERY RECOVERY' l
l FA1 di POWDE R PREPARATION l
FAI PRESSING t
I DEWAXING l
FA2 SINTE RING T
l FA3 GRINDING T
I I
BULK ORYING I
PELLET STACKING r
l FA3 CIRCALOY ROOS ROD LOADING 2
l l
.vE LDING 1i l
STOR AG E i
QU ALITY ASSUR ANCE
.NSPECTION MINT sLEL3ONDL5 KA slLTE RED AIR HEsa,.
ASSE98LY R ADIC ACTIV E AIRSC ANE l
AELEASE POINT SOURCE r
FA1.sA4 REPRESENT v ARIOUS S*ACKS RCCF SHIPVENT MOUNTED ON BLDG.17
- LNTE9ED UO, MWOER Fig. 2.4 Typical block diagram for fuel bundle assembly for the nuclear fuel facility plant, Suilding 17.
Source:
G. Johnstone, CE, personal communication to John WiTliams, July 24, 1981.
l l
2-9 2.5.7.2 Chemistry laboratory The chemistry laboratory performs various quality control analyses on virgin powder and pellets at various production stages to maintain high yields for the fuel fabrication facility.
The pellets are ground and separated for chemical analyses.
Then the material is weighed and dissolved or converted to U 0s, depending on the analyses to be per-3 formed. These laboratory analyses indicate both product quality and concentration of impurities.
2.5.7.3 Metalloorachy laboratory The metallography laboratory receives completed UOz pellets from Build-ing 17 and from the ceramic laboratory for evaluation.
These pellets are sectioned and processed for metallographic examination, which includes tests for porosity and grain size.
2.5.7.4 Fuel and material laboratory This laboratory conducts research and evaluation of nonirradiated core materials that include fuel cladding, neutron poison, and fuel rod assembly materials.
The materials are examined metallographically, mechanically, and chemically for various structural, physical, and chemical parameters respectively.
2.5.8 Radiolooical waste treatment Liquid radioactive waste is generated from fabrication and testing of low-enriched uranium (<4.1".
35U) into fuel assemblies from Buildings 5 and 17.
This waste from Building 17 operations is generated as a result of UO: powder handling, which is carried out in connection with pelleti:ing and rod loading.
The liquid waste consists of floor mop water, cleanuo water, and effluents from sinks and showers in contam-inated (or potentially contaminated) locker rocms used in decontamina-tion operatices.
Radioactive liquid waste from Building 5 laboratories is generated as a result of conducting wet chemical analysis and cleaning of glassware used in the analysis of U0.
In addition, cleanuo water 2
is generated in connection with laboratory fabrication of samole UO:
fuel pellets.
Radiological liquid waste flows to the radiological waste-water treatment plant (8uilding 6) from Buildings 5 and 17.
~~
Radioactive treatment in Building 6 consists of ten 7.5-m3 (2000-gal) storage tanks and four 18.9-m3 (5000-gal) dilution tanks.
All 14 liquid tanks have agitators and recirculation pumos for providing representative samoles.
Based on data provided oy the acclicant, accroximately 4.9 m:
(1300 gal) Der operating day of radioactive waste are oracessed anc discnarged from tne radioactive waste treatment facility.
Building 5 disenarges 3.0 n3 (SCO gal) cer cay, anc Suilding 17 discharges 1.9 m3 (500 gai) cer day of this tctal amcunt of icw-level radioactive uaste.
2-10 The radioactive Wastewater is sampled and analyzed for gross alpha and beta activity before being discharged to an onsite creek and then to Farmington River to ensure levels are below federal standards of 3 x 10-5 Ci/ml (10 CFR Part 20).
2.5.9 Industrial wastewater treatment Combustion Engineering has continued to meet all discharge parameters for industrial wastewater at the Windsor site.
The only sources are noncontact equipment cooling water and demineralized backwash water.
The average discharge flow from Building 17 is 14 m3/d (3700 gpd).
This is composited, sampled, and analyzed for total suspended solids, tem-perature, pH, color, and visible oil sheen, foam, or floating solids in accordance with the National Pollutant Discharge Elimination System (NPDES No. CT000053).
No treatment is required on the industrial wastewater before discharging because industrial waste from Building 17 is not changed significantly in either chemical or physical parameters.
However, in case the indus-trial wastewater is in noncompliance with the NPDES permit, the waste-water will be diverted and stored at the Building 10 area until correc-tive action is taken.
No chemicals are disposed of in the industrial wastewater except caustic and detergent, and, therefore, normally no treatment is required.
Table 2.1 is a summary of CE industrial drain discharge data for 1980.
Table 2.1. Summary of Combustion Engineering industrial effluent data for 198o a
ea Parameter monitored' Type of samate Total suspenced souds 24-n comoosite
< 1o com
<6.oopm Temocrature 4-n average lo*C - 29*C oC 1so F-84'F) i67.7*F) low hign pH 24-h comoosite 6.0 - 8. 3 6.9 tow nign
'samose frecuency is weekly for ndustnal nastewater. A pH oroce mone-tors the effluent uaste continuously and records the data on a stDD Cnart.
source: Data
- rom Combustion Engineering Enwonmentat Imoact Inrorma-tron. Docket No. 70- 1100. Acril 1981 Table 3-3 iRevisedL
{
2-11 2.5.10 Sanitary wastewater treatment Sanitary wastes are generated from Buildings 5 and 17 (radioactive and nonradioactive areas) washroom, locker rooms, lunchroom, and laboratory drains that flow in sanitary waste lines to the sewage treatment plant (Building 10).
All sanitary wastes are processed at Building 10 from nonradioactive areas of Buildings 5 and 17, but only sanitary wastes containing solids are processed from radioactive areas.
The CE sanitary waste treatment method consists of primary and secondary treatment with disinfection of waste by chlorination.
Figure 2.5 shows the sewage treatment plant layout with the various treatment points indicated; there is an average discharge of 11.4 m3 (3000 gal) per day.
Es-tese A
CPW petNGTCM "'yi p
AP8RCXIVATE LCCATION CMSUSTION E NEER NG SITE CREEK.
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9th +0 e Nd SENAGE PLANT CCM8 LEX SECONDARY v
]SEDtMENTAT:CN 00MeuSTICN
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TAN K ENGiNEERtNG PLANT SITE SCUNCARYwl
/ *SANeTARv
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ASTE L;NE l
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.' s S A ces A e7 Fig. 2.5.
Sanitary waste treatment flow sheet, Building 10, serving Buildings 5 and 17.
The treated sanitary waste leaves the system by gravitational ficw over a mari-mace weir and is discharged into the onsite creek.
2-12 2.6 WASTE CONFINEMENT AND EFFLUENT CONTROL All of the radioactive waste generated at the Windsor site originates from either the fuel fabrication facility or the nuclear laboratories.
The total exhaust gas flow averages about 1471 m3/ min (52,000 ft / min) 3 from Buildings 5 and 17.
All of the exhaust gases are discharged via ten stacks, all mounted on the roof.
Of this amount, 1101 m / min 3
3 3
(39,000 ft / min) is from the fuel fabrication facility and 371 m / min 3
(13,000 ft / min) from the nuclear laboratories.
Buildings 5 and 17 will utilize and discharge 38 m3 (10,000 gal) of wastewatef daily ~ from ~~
the operations and support services.
The major process use is for equipment process cooling iri the fuel fabrication facility of 20.8 m3/d (5500 gpd), 5.68 m3/d (1500 gpd) used for other process operations, and 11.4 m3/d (3000 gpd) used for sanitary purposes.
All spent acid and/or hazardous chemicals are collected in outside storage tanks and collected by a local EPA-approved, licensed commercial service on an as-needed basis.
Uranium-contaminated solid waste is collected, nondestructively analyzed (NDA), and disposed of either by recovery or by shallow land disposal.
If sufficient uranium is present, the waste is drummed and trucked to the CE Hematite plant for recovery of uranium after incineration.
Otherwise, the low-level solid waste is shipped to an NRC-licensed, low-level waste disposal site.
According to historical data for Buildings 5 and 17, the radioactive solid waste amounts to 58 m3 (2050 ft ) annually without volume reduc-3 tion.
The bulk of noncontaminated solid waste is collected and disposed of by a licensed commercial firm, and the remainder is collected as trash.
The effluents may contain small quantities of the radioisotopes 23'U, 235U, 23su, and 23eU.
The composition of the mixture will vary deoending on the enrichment of the material being processed; however, in all cases, the bulk of the material will be 23eu (approximately 95% or more by weight), whereas the predominant activity will be 23'U (uo to approx-imately 86% of the total activity).
The average mixture given in Table 2.2 has been used for radiological assessment.-
l The staff has determined that the Windsor fuel fabrication facility meets feceral regulations as set forth in 10 CFR Part 190, 9:vircreenza;
.=adia:icn.=rc:ac:icn handard 'cr.he aar =xer ^rera:icna. 5 (See Accendix A.)
In essence, the regulations require that radioactivity in planned effluent releases from fuel cycle plants be limited so that no member of the public will receive an annual cumulative dose equivalent of more than 25 millirems to tne whole body, 75 millirems to the thyroid, or 25 millirems to any other organ.
2-13 Table 2.2. Average isotopic mixture isotope Mass fraction Actmty fraction Actwity
(%)
(%)
(yCO 22'U o.04 85.38 2.48 22su 4.15 3.07 o.09 23au o.025 o.55 c.16 23su 95.78 11.00 o.32 Source: " Final ~ Enwonmental Impact Apprassal of the Westenghouse Nuclear Fuel Columboa Site. Aprd 1977 Table 3.1.
2.6.1 Gaseous effluent 2.6.1.1 Ventilation systems - radioactive material area Operations involving the use of radioactive materials in unsealed physical forms are limited to low-enriched (<4.1". 2350) uranium in the fuel manufacturing facility or nuclear laboratories facility.
The ventilation systems installed in these facilities are designed to remove essentially all uranium material before final release to the atmosphere from these facilities.
The process ventilation system from Building 17 consists of prefilters, absolute (HEPA) filters, sampling probes, and roof-mounted stacks.
The absolute filter efficiency is certified by manufacturers at 99.97'.'
removal of all discharged particles >0.3 microns.
The radioactive ventilation system for the nuclear laboratories is identical to that of Building 17, except the laboratories do not have prefilters.
Table 2.3 l
shows the estimated airborne uranium releases via the ten stacks.
All exhausts from stacks involving the release of radioactive material are continuously monitored, isokinetically samoled 100". of the time, and analyzed daily.
Stainless steel stack-sampling probes are designed in accordance with American National Standards Institute requirements, as set forth in ANSI N-13.1-1969, and velocity measurements are taken to ensure isokinetic sampling of all ventilation systems.
The process ventilation systems are designed and maintained at a negative pressure in enclosures that maintain an air flow of 100 linear feet per minute (minimum) through the enclosure opening during normal operations.
To ensure no unintentional release of airborne radioactivity to the atmosonere, a negative cressure is maintained inside Buildings 5 and 17.
Exnaust gases are crocerly HEPA-filtered and monitored before discharging.
i
2-14
)
Table 2.3. Estimated airborne uranium releases uent rWease point nu ber
/s)
Bldg. No. 5' 2
Hot chemistry laboratory 1.99 X to-e 3
Emissen spectroscopy 1.28 X lo-e 4
Hydrogen sintering furnace exhaust 5.38 X lo-a 5
Radiochemistry service laboratory, 1.67 X lo-a SNM vault. and environmental laboratory 6
Ceramic laboratory 1.34 X 10-8 7
Metallography laboratory 1.28 X 10-8 and ceramic laboratory Bldg. No.17 FA-1 Powder preparation and pressing 5.01 X lo
FA-2 Furnace H burnoff 1.6o X to
FA-3 Peilet gnnding and rod loading 2.41 X lo~'
FA-4 Recycle powder area 6.15 X 10-8
" Stack No.1 is out of commission and is not being used: Stack No.
8 is presently not operating using radioactive matenal.
2.6.1.2 Nonradioactive process cases The airborne (nonradioactive) chemical effluents are from the pickling operations, from cleaning operations with volatile solvents, and from residual irert gases used in production welding operations.
Exhaust air from these occupied plant areas, except for that removed through equip-ment ventilation systems (discussed in Sect. 2.6.1.1), receives no
.. eatment and is exhausted to the atmosphere by the building roof venti-lation system.
Fumes from acetone, isopropyl alcohol, and Freon are liberated by evaporation during solvent cleaning operations.
Fumes from perchlorethylene will be liberated by evaporation during component and assembly degreasing ooerations.
These materials are not soecifically covered by Federal Air Quality Standards, but the Connecticut CEP iden-tifies these organic chemicals which are photochemically reactive and limits discharges as specified in Table 2.J.
l Table 2.4. Discharge limits - process gases Hourly rate 'imit Daily rate !imit Organic cnemical Acetone, isoproovi alcohol, and Freon 73 16o 36o Soo Perenforetnylene 3.6 3
18 Jo
2-15 Helium, argon, and nitrogen are used in various operational processes in the fuel fabrication plant and nuclear laboratories at a total discharge rate of 85 m3/d (3000 ft3/d).
The gases are discharged through properly located roof exhausts and are sufficiently dispersed at discharge.
Hydrofluoric and nitric acid mists are given off from acid-pickling operations from Building 17 via an exhaust system operating at 140 m3/ min 3
(5000 ft / min). Also, fumes from the aluminum nitrate stop batch are discharged into the same exhaust system, which is operated about 8 h each day.
The average concentration of total nitrogen oxides in the stack from the pickling operation is 9 ppm, and the concentration of hydrofluoric fumes is essentially undetectable (<1 ppm).
No monitoring or treatment of acid mist is required.1 The applicant states in Sect. 4.1.5.4 of Cce uscion Ingineering Inviron-mentad 2:pacc Infomacion:
Because of the small volume of pickling performed and since the concentrations of both N0x and HF fumes are so small, no attempt has been made to determine actual concentration of N0x at the site boundary or away from the C-E site.
It is fei t that actual concentrations, either on or off site, are much less than the national primary and secondary ambient air quality standards for nitrogen dioxide as set forth by the Environ-mental Protection Agency.
The staff's opinion is that the applicant's compliance with air quality standards is ensured.
2.6.2 Liouid effluent 2.6.2.1 Radiolooical wastewater effluent Building 17 generates contaminated liquid waste from floor mop water, cleanup water, and water from the sinks and showers in the change rooms as a result of UO: powder handling, which is carried out in connection with celletizing and rod-loading operations.
Building 5 generates contaminated liquid waste from wet chemical analysis and cleaning of glassware used in the analysis of UO:.
Radiological liouid waste is generated from Buildings 5 and 17 at a rate of J.92 m3/d (1300 gpd),
wnich is diluted to 3.0 x 10-5 Ci/ml (if needed), then transferred to storage tanks at the radiological waste treatment facility (Building 6).
The radiological waste treatment facility consists of ten 7.6-m3 (2000-gal) retention tanks that are automatically filled in sequence.
If eignt tanks are filled to capacity, a blinking warning light located on the cutside wail of the building is activated, wnich indicates that two reserve tanks remain to receive radicactive licuid waste cef:re overflowing.
Overflow water is diverted to a sumo cit.
The acclicant
.v i i i rout nely crocess liquid waste frem two tc tnree tanks for adcec
2-16 safety precaution.
Treatment consists of agitation, circulation, and dilution (if needed) to provide representative sampling.
A 500-m1 sample is withdrawn before discharge and forwarded to the radiochemistry laboratory for gross alpha and beta analysis.
If levels are in excess of 10% MPCw (maximum permissible concentration of wastes in water), the waste liquid is pumped to one of four 18.9-m3(5000-gal) dilution tanks for diluting low-level waste to ensure compliance with 10 CFR Part 20.
Liquid wastes are discharged from Building 6 to the industrial waste line and carried to the CE onsite creek, where they are finally dis-charged.
The total annual release of radioactivity in the liquid waste is about 1.6 x 10-3 Ci from the CE plant."
2.6.2.2 Ncaradiological wastewater effluent Nonradiological liquid effluents are segregated from the CE plant accord-ing to industrial, sanitary, or liquid chemical waste as discussed below.
Sources of nonradioactive liquid sanitary wastes are showers, toilets, sinks, lavatories, and drinking fountains.
Sources of nonradioactive liquid chemical wastes are solvents used to clean components and assem-blies, acid solutions used to clean components and remove residual oxide scale, and degreasing chemicals to remove grease and films from components and assemblies.
Industrial wastewater effluent from Buildings 5 and 17 The details for treatment, disposal, and regulatory control were discussed in Sect. 2.5.9.
Sanitary wastewater effluent from Buildinas 5 and 17 The details for sanitary wastewater treatment and disposal were discussed in Sect. 2.5.10.
Sanitary waste from Buildings 5 and 17 (as well as from the rest of the CE plant) is also monitored and controlled to ensure its compliance with Connecticut DEP regulations.
l The average annual biological oxygen demand (800) for removal of sanitary waste is 90% or higher." Total suspended solids are checked, as are total anc fecal coliform.
As a final disinfectant measure, a free-chlorine residue is maintained in the effluent stream.
A monthly report i
is required and submitted to the stata.
l I
Solids collected from the crimary and secondary settling tanks are routed to tne digester, wnere bacterial action is al cwed to taxe olace and tur*her stacili:e the slucge.
ihe slucge residue is removec tram the digester and spread on sand tecs to dry.
The dried residue is broken uo, removed, and disposed of at a remote location On CE crocerty.
2-17 Liquid chemical waste effluent from Buildings 5 and 17 The various quantities of liquid chemicals used are listed in Table 2.5 for disposal of liquid waste from the CE site.1>5 All spent acids and organics (not evaporated) are disposed of from the plant as needed by an EPA-state-approved commercial pickup service in Waterbury, Connecticut (G. Johnstone, CE, personal comunication to John Williams, July 24, 1981).
All organics from various processing operations (e.g., cleaning) are evaporated naturally or in furnaces.
The remaining chemicals are disposed of via the industrial waste treatment circuit.
2.6.3 Solid waste effluent 2.6.3.1 Radiological solid waste effluent Solid wastes containing UO2 are generated in Building 17 in the form of rags, paper, plastic liners, prefilters, rejected empty fuel tubes, and other miscellaneous materials generated during normal processing opera-tions.
Radioactive solid waste is generated in Building 5 in the form of paper, rags, poly bags, and other refuse that bears precipitated uranium from chemical processes and analyses in solid form.
Radioactive solid waste is also generated in the form of scrapped, unusable 00 powder in small quantities.
All waste materials are loaded into either a wooden burial box or 0.21-m3 (55-gal) drums and placed in the Bulk Assay Counter to determine the 2350 content by NDA.
The drums are sealed, labeled, and shipped to either an NRC-licensed low-level waste disposal site or to the CE Hematite plant for recovery of 235U by incin-eration and further processing.
The used absolute filters are nondestructively assayed to determine the amount of 235U contained in each filter.
They are then either buried or disassembled to recover the uranium.
A representative sample is taken for uranium analysis for recycle purposes.
In 1980, the solid waste generated and disposed of was approximately 79 kg (174 lb) of uranium, or2.3kg)(ofsolidwaste.
5.1 lb) of 2350 from Suilding 17.
This reoresented 36.3 m3 (1700 ft l
l 2.6.3.2 Nonradiological solid waste effluent The bulk of the ncnradioactive solid wastes is collected and discosed af by licensed commercial firms.
Other valuable scrap metals are segregated by type and ver1tled as containing no detectable uranium.
Then they are either returned to the manufacturer for reorccessing or are sold to scrap dealers for eventual recycling.
Ola fixtures, tools, and ecuip-ment are diseased of to ccmmercial scrao dealers.
Table 2.5. Chemicals used at the fasel fabrication plant Arviuol disclarge Despersal Cliumn:ot LeninAs means L
9W kg h
At eninie Cl1,C -- OCl1, 2.271 600 Evaporation lequigeyl ol..oluil CH,Ca lOllCit, 4,164 1.100 Evaporation liphinlil<n o. scal 18C1 11.025 5.000 lewiustrial waste ettksent iIyihorlinu o. soul iIF 110 50 Licensed commercoal pickup service Ahununun suuate ANtJO),
2.646 1.200 Licensed conunercial pckup service
~
Sininun liyih umules tJaOli 3.969 1.800 lodustnal waste effluent iJ iro. tui
- lirJO, 3.028 800 Licensed commercial pckon service Deter 9ent sointuu.
1.984 900 industnal wasta elftuent to a
Peettilineth tene Cit -CitCl 24.696 11,200 Evaporation i
tseon CCL),
243 110 Evaporaison Mai huie a oulant 243 110 Licensed conunercial pckup service vacuiun launn ud 757 200 Licensed conunercial pckup servce Polyvuigl a6cohol CitOiICli CitOllCil, 13.230 6.000 Evaporaison
/nic steasate C,,lt, COO Zn 2.866 1.300 Evaporation Atutnuntuso pn klaaj liath litJO, i itF l 11,0 3.785 1.000 Licensed commercial pckup servce Sinus u. I'sunanly taken tsnun Cannbuston Etyuiceruy Envsrorunental Inquet Information. Sect. 4.1.4.3. Lutuid Chenucal Waste i illuent t
t W
1 2-19 2.7 DECOMMISSIONING At the end of its orarating life, the plant will be decontaminated and decommissioned before the site and any plant buildings remaining on the site can be released for unrestricted use.
In accordance with NRC requirements, the applicant has prepared and submitted a decommissioning plan, cost estimate, and financial surety, dated January 12,1979, for inclusion in the license application materials.
The major guidelines embodied in the plan are as follows.
1.
All buildings are to be cleaned to levels established for unrestricted use.
2.
Current radiological limits and decontamination technology are to be utilized.
3.
All process and ancillary equipment in controlled areas is to be cleaned to the extant practicable, packaged, and transported to a licensed disposal facility for burial.
4.
Any contaminated underground piping is to be removed, cleaned to the extent practicable, packaged, and transported to a licensed disposal facility for burial.
The ground surrounding such piping is also to be surveyed and removed for disposal if contaminated beyond established limits.
5.
Packaging, transportation, and disposal charges are to be calculated using information from existing licensed low-level waste disposal facilities.
The NRC has prepared a two-volume report to provide information on the technology, safety, and costs of decommissioning uranium fuel fabrication plants.
This information is intended to contribute background data for uranium fuel fabrication plant owners and for the NRC and to provide the basis for_ future regulation regarding decommissioning of such facilities.
2.8 MATERIAL CCNTROL AND ACCOUNTABILITY Current safeguards are set forth in 10 CFR Parts 70 anc 73.
The regula, tions in Part 70 provide for material accounting and control requirements with respect to facility organization, material control arrangements, I
accountability measurements, statistical controls, inventory metncds, j
shipping and receiving procedures, material stcrage oractices, records and reports, and management control.
l l
The Ccmmission's current regulations in 10 CFR Part 73 orovice require-ments for the anysical security anc arotection of fixed sites and for j
nuclear material in transit.
Physical crotecticn reauirements for 3:ecia!
l nuclear naterial of low strategic significance (inc!ucing icw-enricnec l
uranium; incluce crovisicri 0r estaoiiscrent o' controllec access areas.
l l
l
o 2-20 monitoring these areas to detect unauthorized penetration, providing a response capability for unauthorized penetrations and activities, and establishing procedures for threats of theft and thefts.
The Commission's regulations in 10 CFR Parts 70 and 73, described briefly above, are applied in the reviews of individual license applications.
License conditions then are developed and imposed which translate the regulations into specific requirements and limitations that are tailored to fit the particular type of plant or facility involved.
The licensee has an approved material control and accounting plan and an~
approved physical security plan which meet the current requirements for the low-enriched uranium which would be possessed at the site.
It is concluded, therefore, that the safeguards-related environmental impact of the proposed action is insignificant.
2.9 STAFF EVALUATION OF THE PROPOSED ACTION AND ALTERNATIVES The staff has concluded that the denial of license renewal would provide very little in the way of environmental benefits.
The staff believes that the fuel fabricating processes used at this facility are efficient and that the methods of waste confinement and efficient controls meet all applicable state and federal standards. The staff recommends that Special Nuclear Material License No. SNM-1067 be renewed.
2-21 REFERENCES FOR SECTION 2 1.
Combustion Engineering Environmental Impact Information, Combustion Engineering, Inc., Windsor, Connecticut, USNRC Document No. 70-1100, April 1981.
2.
U.S. Nuclear Regulatory Comission, Environmenect Impact Appraisal of the Westinghouse Nuclear hal Columbia Site (NFCS) Commercial Nuclear Fuel Fabrication Plant, Columbia, South Carolina, Report NR-FM-013, April 1977.
3.
U.S. Code of Federal Regulations, Title 40, Part 190, " Environ-mental Radiation Protection Standards for Nuclear Power Operations,"
Superintendent of Documerits, GPO, Washington, D.C. 20555, Rev.
July 1, 1980.
4.
Combustion Engineering, Inc., questions and Responses Related to Nuclear Fuel Fabrkation Facility and Nuclear Laboratories Envi-ronmental Impact Informa: ion, Response 1981, Docket No. 70-1100.
5.
Environmental Impact Appraisal of Sabcock & Wilco Material Division, Comercial Nuclear Fuel Fabrication Plant, Borough of Apollo, Pa., October 1978.
6.
H. K. Elder and D. E. Blahnik, Technology Safety and Costs of Decommissioning a Reference Uranium ?uel Fabrka: ion Plant.
Report of U.S. Nuclear Regulatory Comission by Pacific Northwest Laboratory, NUREG/CR-1266, Vol.1, October 1980.
7.
H. K. E1 der and D. E. Blahnik, Technology Safety and Costs of Deconrrissioning a Reference Uranium Fuel Fabrica ion Plan:.
Report of U.S. Nuclear Regulatory Comission by Pacific Northwest Laboratory, NUREG/CR-1266, Vol. 2, October 1980.
I
3.
AFFECTED ENVIRONMENT 3.1 SITE DESCRIPTION The CE Windsor site is a 225-ha (556-acre) tract located in the township of Windsor, Connecticut (Fig. 3.1).
It is 14.5 km (9 miles) north of Hartford and 4.8 km (3 miles) south of East Granby, the nearest town.
The town of Windsor is about 8 km (5 miles) southeast of the site.
Buildings 5 and 17 are the ones primarily involved in the activities subject to the proposed relicensing. Coordinates for Building 17 on the.
site are N 41 52'56" and W 72*43'08".
About 7% of the site is developed with buildings, parking lots, roads, etc. (ER, Fig. 2-9).
The remainder consists primarily of woodland.
The dominant surficial features are primarily a result of glaciation during the Pleistocene epoch and subsequent erosion.
Site topography is level to rolling with little relief.
3.2 CLIMATOLOGY AND METEOROLOGY 1 Climatologic 7' and meteorological data are based mainly on measurements made by the Weather Bureau Station located at Bradley International
.a.
Airport, aoout 8 km (5 miles) northeast of the plant site.
The mean temperature for the 50-year period 1931 through 1979 was 10 C (50 F), as recorded at the Bradley International Airport.
The maximum and minimum monthly mean temperatures were 28.4 C (83.2 F) and -12.2 C (10.1 F) respectively.
3.2.1 Winds, includino atmosoberic discersion The highest recorded wind velocity was 31 m/s (70 mph).
The crevailing wind direction for the six months May through October is from the south; for the six months November through Acril, it is from tha northwest.
The average wind velocity at the plant site is 5 m/s (11.2 mph).
With low-to-moderate wind speeds, inversion conditions may exist from sunset to sunrise. A strong laose rate exists around noon; the temaer-ature d1trerence is maximum with air flow upward at a maximum rate.
As the night approaches, weak lapse conditions occur with low air flow.
There are significant continental and oceanic influences on the area's meteorological and climatological conditions.
With tne crevailing west-to-east air flow, continental modifications of :ne air are imcortant.
However, sudden and oftentimes serious uosets result unen storms move north or when otner cressure develoaments ::recuce the streng anc cer-sistent nortneast aincs asscciated witn storms kncwn locally as 'ccastals" or 'nortneasters." Seasonaole air mass cnaracteristics vary frem the 2-1
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extremely cold and dry continental polar quality of winter to the warm, humid, maritime, tropical characteristics of summer.
Atmospheric dispersion models and parameters are discussed in Appendix A.
3.2.2 Precipitation Local topography also influences the climate.
The Berkshire Hills to the west and northwest are a source of summer thunderstcrms that may be accompanied by wind and hail.
Frequently during winter, rain falls tnrough the cold air trapped in the valley with resultant icing condi-tions.
On clear nignts in the late summer and early autumn, cool air drainage into tne valley, plus Connecticut River moisture, produces ground fog that scmetimes becomes ouite dense.
3-3 The total precipitation for 1979 was 97.6 cm (38.4 in.), with a maximum of 13.1 cm (5.2 in.) falling in February).The mean annual precipitation from 1931 to 1979 was 107.5 cm (42.3 in..
The maximum monthly precipi-tation was 55.6 cm (21.9 in.) in August 1955.
3.2.3 Severe weather The flood level for the area's worst flood (August 1955) was about 33.5 m (110 ft),above mean sea level.
Because the plant site is located approxi-mately 54.9 m (180 ft) above mean sea level, the probability of direct damage resulting from local flood waters is considered very low.
Tornado and straight wind probability hazards have been calculated in a detailed report prepared by the Institute of Disaster Research2 for an area of Connecticut that includes the Windsor plant site.
The tornado hazard probability for this region was estimated to be 1.0 x 10-6 for a tornado with winds ranging from 62.6 m/s (140 mph) to 106.8 m/s (239 mph), or one chance in one million of a tornado of this size affecting the site.
3.3 AIR QUALITY The CE Windsor site is located in the Hartford-New Haven-Springfield Interstate Air Quality Control Region.
In this region, concentrations of particulates do not meet national secondary standards, and concen-trations of ozone and carbon monoxide do not meet primary standards (40 CFR Part 81, revised July 1, 1981).
Concentrations of sulfur dioxide and nitrogen dioxide are below the standards.
Neither ambient air quality nor emissions to the atmosphere are monitored for nonradiologic pollutants at the Windsor site (ER, Sect. 5.1.2.2).
The staff obtained air cuality data for the Hartford and Enfield stations near the site (Table 3.1).
Air cuality is generally poorer at the more urban Hartford station than at the Enfield station.
Air quality at the Windsor site is probably intermediate between that at the Hartford and Enfield stations.
The concentration of ozone at the Hartford station l
exceeded the national ambient air quality standards for chotochemical l
oxidants (Table 3.1) (ozone data are not collected at the Enfield station).
l Concentrations of S0x, N0x, and particulates at the two stations did not exceed any of the standards.
Ambient particulate concentrations at the j
two stations range from 50 to 90% of the national standards, S0x con-centrations range from 22 to 74% of the standards, and N0x concentraticns range from 50 to 80% of the standards.
No Class I Prevention of Signifi-cant Deterioration areas are located near the site or in tne state of Connecticut.
I l
O F
3-4 Table 3.1. Summary of air quality data for the Combustion Engineering Windsor site Primary standards were developed to protect human health.
Secondary standards were developed to protect environmental values.
Parameter National air Connecteut oepartment of Environmental Quahty data 3
fugim )
quahty standard Pnmary Secondary Enfield*
Hartfield#
Sulfur dioxxte 24 h 365 26o 105 (81f 200 (193)
Annual average 80 60 13 38 d
TSP 24 h 26o 15o 82 (78) 167(109)
Annual average 75 60 36 54 Nitrogen dioxide 24 h 113 (104) 157 (154)
Annual average too loo 51 79 Ozone Ih 200 200 53o (480)
'Enfield is 13 km (8 miles) northeast of the Combustion Engineering Windsor site.
8 Hartfield is 16 km (10 miles) south of the site.
- Highest measured concentrations and second highest concentrations (in parentheses).
- Total suspended particulates.
~
3.4 SOCI0 ECONOMICS The area surrounding the CE Windsor site is sparsely populated.
East Granby, the nearest town (Fig. 3.1), is located approximately 4.8 km (3 miles) north of the site.
It has a population of 4039 persons, yielding a population density of 591 per square kilometer (228 per square mile).
Windsor is the nearest large town.
Its town center is l
located approximately 8 km (5 miles) southeast of the site.
The town's population is 25,171 persons, yielding a population density of 328 cer square kilometer (850 per square mile).
The estimated peculation l
within 80 km (50 miles) of the site is 3,328,490 persons.
The distri-bution is shown in Table A.8 in Appendix A.
The above data is based on preliminary 1980 census data.
I 3.5 LAND USE i
The CE Windsor site is located in the township of Windsor, which covers about 7795 ha (19,261 acres) (ER, Sect. 2.1.3).
A summary of land use in Windsor is presented in Table 3.2.
The amount of land occupied by residential and commercial areas and public facilities is increasing at l
the excense of agricultural and vacant lancs.
Residential areas, oublic facilities, and agricultural areas each occupy about one-fourth of the total area.
About 555 of the agricultural land is used for growing
3-5 Table 3.2. Approximate percentage land usu in the townships of Windsor, Bloomfield, and East Granby Windsor 1977 1979 Bloomfield East Granby Residential 24.0 24.7 15.3 8.7 Commercialhndustnal 7.3 7.7 5.8 2.2 Pubhc facilities 14.5 14.8 28.0 10.4 Agnculture 25.3 24.7 9.1 30.9' W ater 5.6 5.6 1.2 0.0 Vacant 23.1 22.4 40.6 47.8*
Total area (ha) 7794.8 7794.8 6967.2 45o6.7
'Agnculture and open space.
- Woodland.
Source: ER. Tables 2-3. 2-4, and 2-5.
broadleaf tobacco.
The nearby townships of Bloomfield and East Granby have a considerably greater fraction of vacant land, primarily woodland (Table 3.2), than does Windsor.
On the site, there are 27 buildings (ER, Table 2-6 and Fig. 2-7), occupy-ing 7.1 ha (17.5 acres) of the 225-ha (556-acre) site (ER, Sect. 2.1.4).
An additional 8.9 ha (22 acres) is devoted to access roads, parking lots, walkways, etc.
Thus, 7% of the site is developed.
The remaining 93% consists mainly of woodlands.
Located in the immediate vicinity of the site are the Farmington River and a town-owned sanitary landfill to the north; wooded areas, agri-cultural fields, and the Griffin Office Center to the south and west; and a federal facility, the Knolls Atomic Power Laboratory, to the east.
Three significant industrial complexes exist within an 8-km (5-mile) radius of the site.
The Hamilton Standard complex is in Windsor Locks and emoloys 7000 persons.
Stanadyne, Inc., and Emnart Industries in Windsor employ 3000 and 720 persons respectively (ER, Sect. 2.1.3).
Recently, two additional land parcels totaling 223 ha (550 acres) have been added to the site.
One is a 200-ha (493-acre) tract to the soutn that is mostly agricultural.
No plans have been made for this tract, and the agricultural use will continue in the near future.
A 23-ha (57-acre) tract to the west is primarily a wooded area and a gravei cit.
No plans have been made for this area (ER, Sect. 2.1.5).
Windsor's historic survey (1981) lists no historic sites in the vicinity of the CE site (ER, Accendix O.
3-6 3.6 WATER 3.6.1 Surface water The major water bodies in the Windsor area are the Farmington and Connecticut rivers.
The Farmington River flow approximately 1.2 km (0.75 mile) north of the CE site.
A small c eek draining the site empties into the Farmington River upstream of the Rainbow Fishway at the Farmington River Power Company's Rainbow Reservoir Dam.
The Farmington River then empties into the Connecticut River approximately 16 km (10 miles) downstream of the CE site.
Both rivers are used primarily for recreational boating and fishing.
3.6.1.1 Hydrolocy The flow of the Farmington River in the vicinity of the CE site is regu-lated by the Rainbow Reservoir Dam and nine upstream reservoirs. and dams as well as by diversions for municipal supplies.
The 50-year average discharge at the U.S. Geological Survey Rainbow Dam gaging station downstream of the CE site is 30.78 m3/s (1090 ft /s) with a maximum 3
discharge of 1960 m3/s (69,200 ft /s).
The maximum gage height was 3
7.2 m (24 ft) on August 19, 1955, and the minimum discharge was 0.14 m3/s 3
(5 ft /s) on four occasions during the 1940s.3 3.6.1.2 Water quality The Farmington River, which receives discharge from the site, is classified as suitable for fish propagation and is protected for propagation of the Atlantic salmon (Jim Moulton, Connecticut Fish and Wildlife Unit, Hartford, Conn., personal communication to V. R. Tolbert, Oct. 14,1981).
In the vicinity of the CE site, the river is rated as a Class B Inland Water and is considered suitable for " bathing and other recreational purposes, agricultural uses, certain industrial processes and cooling; excellent fish and wildlife habitat; aesthetic value; acceptable for public water supply with aporopriate treatment."" General water quality characteristics in the vicinity of the site are shown in Table 3.3.
3.6.1.3 Water use The Barkhamsted, Neoaug, and West Hartford No. 2 reservoirs supply water for the cities of Hartford, West Hartford, and Windsor.
Barkhamsted Reservoir, the laraest reservoir of the Metropolitan District Commission (MDC), is located over 20 km (12 miles) uostream of the CE Windsor site.1 The major volume of water used at the site is suoplied by the MCC tnrough a 30.5-cm (12-in.) main that enters the southern boundary of the site at Prospect Hill Road.
Additional water used for ccoling purcoses is sucolied by two wells located onsite.
Of a total of 1120 v, (300,000 gal) of water per day delivered to the site by the MCC, only
3-7 Table 3.3 Selected water quality parameters of the Fermington River in the vicinity of the Combustion Engineering Windsor plant site Spenfc pH Dessolved Sulfate Chlorine NH Dissolved Organc 4
conduct.
oxygen (mg/L)
(mg/L)
(mg/L) solids carbon umho (mg/L)
(t/d) fmg/L)66-148 6.5 - 7.3 8-13.4 9-12 7.4-14 0.1-o.4 60-960 2.3-15 Source: U.S. Geologeal Survey.1978.
38 m3 (10,500 gal) per day are directed to Bldgs. 5 and 17, the buildings covered by this license renewal application.
Each of these buildings consumes apprcximately one-half of this total, or 19 m3 (5200 gal) per day.
Figure 3.2 indicates the types and volumes of water utilized for processing, cooling, and sanitary purposes and the discharge from these buildings.
Treatment and discharge of water used in processing is discussed in Sect. 1.2.
3.6.2 Groundwater 3.6.2.1 Hydrology The plant site is underlain by two types of possibly water-bearing materials:
unconsolidated materials primarily of glacial origin and bedrock of the Mesozoic era (see Sect. 3.7.1).
Both of these sources appear to have a limited potential for the development of high-yielding water-supply wells.
3.6.2.2 Groundwater use Although the site presently operates two wells, groundwater drawn from these wells is not used in the fuel fabrication process.
Approximately 38 m3/d (10,000 gpd) of water are supplied to Bldgs. 5 and 17.
Each of these buildings in turn consumes about one-half of the total.
This water is supplied by the MDC, wnicn gets its water from a variety of sources, the largest being the Barkhamsted Reservoir.
3.6.2.3 Grounawater cuality The two wells located on the plant site are sampled quarterly and analy:ed for cri, fluoride, nitrate, uranium, Inc aloha and Deta r3dio-activity levels.
The ranges for these carameters during tne 1975-1979 interval are given below:
e e
3-8 ES-5896 PROCESS INDUSTRIAL
- p# os gd p WATER S
x WASTE WATER I
Y ht p
BLDG.5 EQUIPMENT 6.00 X 10 gai DEVELOPMENT 5
5 1.92 X 10 gag RADIOLOGICAL COOLING DEPARTMENT
=
WASTE WATER LABORATORIES WATER h
0%y
- Agog, y
g, SANITARY WATER FOR WA SANITARY USE R
PROCESS INDUSTRI A L WATER 2
WASTE de* /os d 4*
WATER g9*g B LDG.17 EQUIPMENT 7.20 X 105 gal NCCl EAR FUEL 1.20 X 1@ gal WlOmim COOLING L
M Ar.UF ACTU RING W ATE R WATER h
- os s
3##
WATER FOR IE SANITARY USE q
Fig. 3.2.
Water bala.!Ce diagram for Buildings 5 and 17 on an annual basis (240 working days).
Comoiled frcm responses tO Comments by the applicant.
3-9 pH 7.0-8.4 Fluoride (pp.n)
<0.10-0.12 Nitrate (ppm) 1.1-16.4 Uranium (ppm)
<0.001-0.001 Gross alpha (pCi/L)
<l.0-3.5 Gross beta (pCi/L)
<l.3-16.1 These wells are not used for drinking water purposes.
3.7 GE0 LOGY, MINERAL RESOURCES, AND SEISMICITY 3.7.1 Geoloay A detailed geologic description of the plant site has not been performed; however, two geologic investigations,6 of the property adjacent to or s
near the site have been completed, and the following general geologic description has been drawn from these two studies.
The surficial material on the plant site probably consists of a mixture of unconsolidated material that was deposited by glacier ice or meltwater streams.
Unconsolidated materials that might be present on the site, from oldest to youngest, would consist of:
till that is unsorted, unstratified, and contains a heterogeneous mixture of clay, sand, gravel, and boulders varying widely in size and shape; ice-contact stratified drift consisting of mostly well-sorted sand and gravel; glacial deltaic deposits consisting primarily of sand and fine sand and some gravel and silt; terrace sediments consisting of sand, gravel, and silt that were deposited by the Farmington River on flood plains of earlier origin; alluvial sediments from the modern floodplain of the Farmington River; and swamo sediments consisting of silt, clay, sand, and organic material that have accumulated in shallow depressions.
Bedrock cropping out and underlying the site probably consists mostly of reddish-brown arkosic siltstone with some beds of reddish-brown arkosic sandstone and gray siltstone of the Mesozoic era.
3.7.2 Mineral resources The staff is not aware of the potential for any mineral rescurce that could be actively exploited witnin or near the plant boundaries, with
3-10 the possib!e exception of sand and gravel.
The staff believes that sand and gravel would not be considered a scarce resource in this area and could be actively surface mined in many other locations in the area.
3.7.3 Seismicity The plant site is located in a minor-to-moderate-risk seismic zone (Fig. 3.3).
The staff believes that it is highly unlikely that a large-magnitude earthquake will affect the plant site during its projected life.
~
3.8 ECOLOGY 3.8.1 Terrestrial No vegetation and wildlife surveys have been conducted on the CE Windsor site.
The applicant supplied information on endangered species from the state of Connecticut and on vegetation from surveys at Northwest Park, 2 km northeast of the site.5 The forest vegetation on the site, probably similar to that at Northwest Park, is described in Table 3.4 and Fig. 3.4.
No wildlife surveys were conducted at Northwest Park.
The forests at the site and vicinity probably support more than a hundred species of amphibians, reptiles, breeding birds, and mammals.
Also, many addi-tional species of birds use the forest temporarily during migration.
Information on the animal species that occur in the region can be obtained from several readily available field guides.7-9 3.8.1.1 Endangered species Endangered animal species whose ranges include the CE site are the Indiana bat (Myo:is sedalis), eastern cougar (Felis ccncc!cr acucar),
bald eagle (Ecliaeerus leuccesphalus), and peregrine falcon (Falec peregrinus).
No federally listed endangered plant species occurs in Connecticut.10 The state of Connecticut has a list of rare plant species.
l However, the list is an advisory list and carries no legal status (ER,
[
Appendix A).
Nont af the endangered animal species is expected to occur at the site.
The-appears to be no habitat at the site that would make the area atti..tive to any of these scecies.
Critical habitat for the Indiana bat includes only 13 or so caves where the bats concentrate for hiberna-tion during the winter.ll The cave closest to Connecticut is in West
'/i rgi ni a.
During the summer, Indiana bats are widely dispersed in woodland nabitats; it is unlikely that very many, if any, occur at the site.
The few studies that have been conducted on this scecies indicate
- nat the bat prefers woodlands along small streams. R As of 1976, the
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MtLES O 50 100 150 200 0
10 0 200 300 KILOMETERS Fig. 3.3.
A preliminary map of horizontal acceleration (expressed as percent of gravity) in rock with 90". probability of not being exceeded in 60 years.
Source:
S. T. Algermission and 3. M. Perkins,.4 Probcbi;ia:ia Ea:i.ra:a a;' '::=i.v.c:.4a:. e:e.~ :icn in.=cck in :ha :antipcua "ni:ed 5:a:es, Open File Report 76-426, U.S. Geologic Survey, 1976.
f Table 3.4. Vegetation types of Northwest Park near the Combustion Engeneering Windsor site huest type Sint= asul Ueusnos pluAngy Dommunt tree specses Donunant steub species Structural characsimistics Oak Leo.su:oins Dry. wed diouieal Scasiet ook. pitch pene.
Huckleberry, low bush Ground cover often other oaks. Buckosses, blueberry sparse, dwari siwth cosufers layer Oak Vdiurruun Mesac escargunents and Red o.A. tAack berch.
Maple leafed viburnum.
Sivub layer was siumlerately weR red maple beaked tutobiut, witch developed thauu:d bassis hazel Oak Hed nui e-Pously drauned. near Hed maple Spacebush, arsowwood.
Dense ground cover of d
Cunianumi teen sw,,np3 a,xj t,tooks tughbush tAueberry cannamon forn, open shrub layer s.>
Heil nuple.
Dessessions unautated Red maple, gray tsch Higitush IAueberry.
2-to 3-m-hgh thschet with e
)
Wuatesl>ee ry un speuuj asul early swariip arislea, winterbenry scattered red rnapio N
Atalea sununer hoNy Heil nuple Piukled un sprug and Red maple, buttonbush Lack of trees few Snuu weeil cady statuues starubs, domuuted by Pohgonum (smartweedi S.mps nuple Ferade stuls. nuus Sugar maple, wtute Spacebush. northern Weg developed sivub Wlute a:>h to wet lower slopes ash, sed oak, tuhp arrowwood. witch hazel layer osul vaueys trees 9
3-13 ES$906 A
O G B
A E
A C
C A
F A
B A
F O
5 8
9 e
i E
5 i
9 5
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fsP j,o ySTEM {
TILL MI LANDS LAND SYSTEM -[
TERRACE AND PLAIN $ (T&P) LAND SYSTEM l
Fig. 3.4 Diagrammatic cross section cf Northwest Park showing the typical toposequence of plant comunities in the CE Windsor site area.
A = oak-ericaceous forest, B a oak-viburnum forest, C = oak-cinnamon fern forest, D = red maple-skunk cabbage forest, E = maple-winterberry azalea forest, F = sugar maple-winterberry azalea forest, F = sugar maple-white ash forest, G = open water.
Indiana bat was reported from only a single locality in Connecticut, the habitat of which has since been destroyed.13 The eastern cougar subspecies was extirpated in Connecticut by the early 1800s. During the last several years, many sightings of cougar in the eastern United States have been reported. However, none or very few of these have been verified by specimens, photographs, or casts of tracks.
Also, individuals of the western subspecies that escaped from captivity may be responsible for the sightings. Cougar in general require large expanses of remote wilderness. Therefore, it is highly unlikely that any exist in the vicinity of the site, although some might occur in wilderness areas in western Connecticut.13 The peregrine falcon is extremely rare as a breeding bird in the eastern United States.
It formerly nested on high rocky cliffs in the central lowlands of Connecticut,13 but no cliffs occur near the site.
Currently, the falcon exists only in a few places in the eastern United States where pen-raised individuals have been introduced by Cornell University and the U.S. Fish and Wildlife Service.
In addition, one wild pair recently nested in eastern Maine and fledged two young for the first successful nesting in the area since 1955.1" None of the known nest sites is located near the CE Windsor site. The peregrine falcon occurs more often, but still rarely, as a migrant in the eastern United States.
It woulo not be expected to occur regularly at the site.
The cald eagle occurs as a breeding bird, as a migrant, and as a winter visitor in the eastern United States.
It nests in various localities such as Chesapeake Bay and the Potomac River areas near the Atlantic Ocean. No eagle nesting sites are kncwn to exist in the vicinity of the CE site, and eagles are not known to occur at tne site during migration or winter.
Formerly, eagles nested in a few remote areas of Connecticut.M
3-14 3.8.2 Aouatic Rainbow Reservoir, located on the Fannington River downstream of the CE site, has a surface of 94 ha (234 acres), a maximum depth of 15 m (50 ft), and an average depth of 6 m (19 ft).
Dense growth of aquatic vegetation occurs in most of the shallow areas of the reservoir, and heavy algal blooms have been recorded.5 In the vicinity of the CE site at Rainbow Reservoir, the Farmington River contains abundant populations of several fish species and a number of common fish species (Table 3.5).
In addition to these resident populations, several anadromous species ( Atlantic salmon, American shad, and sea lamprey) migrate via fish ladders up the Connecticut and Farmington rivers as far as 32 km (20 miles) upstream of Windsor.13 A ladderless dam at that point stops their migration.
The American shad spawn and utilize nursery areas throughout the available range in the Farmington River during the summer and early fall before their return to the ocean. Atlantic salmon migrate in sufficient numbers in the spring and fall to provide a sport fishery in the upper Farmington River.
Trout are stocked in the Farmington River in the vicinity of Tariffville Gorge upstream of Windsor. The potential exists for trout to occur in the vicinity of the CE site.
Table 3.5. Fish species occurring in the Farmington River in the vicinity of the Combustion Engineering Windsor site Common name Scientific name Abundant Largemouth bass Micropterus salmorces salmordes l
Red-breasted suntish Lecomes auntus Common suntish Lecomrs gibbasus Bluegini sunfisn Lecomas macrochirus macrochirus Golden shnner Notemogonus chrysoleucas Amerscan eel Anguilla rostrata Common Yetlow peren Perca flavescens Chain pickerei Esor nyer Rock bass Ambicotstes rupestns rupestris Whrte suckers Catostomus commerscno commersono Brown bus \\ head Ictaturus nebulosus l
Smanimoutn bass Microoterus dolomieur dosomeur Migrate ry l
Atlantic salmon Salmo salar
(
American snad Atosa sao,dissima Sea tamorev Petromvron marinus l
l L
3-15 Rainbow Reservoir is generally in a productive state of balance and should provide a good-to-excellent fishery for largemouth bass, chain pickerel, and yellow perch.s Current lack of access causes the reservoir to be underutilized as a fisheries resource.
1 3.9 RADIOLOGICAL CHARACTERISTICS (BACKGROUND)
On the basis of data from Natural Radiation E.moeure in the United States,15 the total-body dose rate from natura'l background radiation in the vicinity of the Windsor plant site is expected to be on the same order as that of the state of Connecticut:
110 millirems / year (41 millirems / year from cosmic radiation, 51 millirems from background terrestrial sources, and 18 millirems / year from naturally occurring radioactive materials deposited in the body).
i i
l t
3-16 REFERENCES FOR SECTION 3 1.
U.S. Nuclear Regulatory Coninission, Environmental Imract Infonnation, Combustion Engineering, Inc., Nuclear Fuel Fabrication Facilities, Nuclear Labaratories, Windsor, Connecticut, Docket No. 70-1100, April 1981.
2.
J. R. Mcdonald, Tornado and Straight Wind Har.ard Probability for Haddam Neck Nuclear Power Reactor Site, ' Connecticut, Institute for Disaster Research, Texas Tech University, Lubbock, Tex., 1980.
3.
U.S. Geological Survey, Water Resources Data for connecticut, U.S.
Geological Survey, Water Data Report CT-78-1 (1978).
4.
W. S. Fuss, Solid Waste Disposal Study, H'<ckleberry Road Site, Town of Windsor, Geraghty and Miller, Inc. (1971).
5.
King's Mark Resource Conservation and Development Area Environmental Review Team, " King's Mark Environmental Review Team Report on Northwest Park, Windsor, Connecticut," Warren, Conn. (December 1980).
6.
Town of Windsor, " Solid Waste Disposal Study, Huckleberry Road Site, Town of Windsor," Windsor, Conn., September 1971.
7.
R. Conant, A Field Guide to Reptiles and Amphibians, Houghton Mifflin Co., Boston, Mass., 1958.
8.
C. S. Robbins, B. Bruun, and H. S. Zim, A Guide to Field Identifi-cation of Sirds of Nor:h America, Golden Press, New York, 1966.
9.
W. H. Burt and R. P. Grossenheider, A Field Guide to the Mcmals, l
Houghton Mifflin Co., Boston, Mass.,1964 I
i 10.
Department of the Interior Fish and Wildlife Service, 50 CFR Part l
17, " Republication of the Lists of Endangered and Threatened Species and Correction of Technical Errors in Final Rules," Fed.
Regist., 4,5(99), Tuesday,May 20, 1980.
11.
50 CFR Part 1.1:109, Revised October 1979.
12.
S. R. Humphrey, A. R. Richter, and J. B. Cope, " Summer Habitat and Ecology of the Endangered Indiana Bat, Myo:is sodalia," J. Mamal.
58: 334-346 (1977).
13.
J. J. Dowhan and R. J. Craig, " Rare and Endangered Species of Connecticut and their Habitat," State Geological and Natural History Survey of Connecticut, Department of Environmental Protec-tien, Report of Investigations No. 6 (1976).
1
3-17
- 14. American Birds, 34_(6): 875(1976).
15.
D. T. Oakley, Natwnt Radiation E=pceure in the United States, ORP/SID 72-1, U.S. Environmental Protection Agency (June 1972).
w
f 4.
ENVIRONMENTAL CONSEQUENCES OF PROPOSED LICENSE RENEWAL 4.1 MONITORING PROGRAMS AND MITIGATORY MEASURES 4.1.1 Onsite monitoring program The onsite monitoring program conducted by the applicant is shown in Table 4.1.
Locations of sampling relative to the plant layout are shown in Fig. 4.1.
Table 4.1.
Onsite environmental monitoring program for the Combustion Engineering Windsor plant' Sample type Frequency Location Air mocitoring Building vents Continuous Bldgs. 5 and 17 (12 vents)
Atmospheric fallout Quarterly At 14 samphng stations around the site Water monitoring Grao samples of surface Quarterly One sample each at the water site creek at Farmington River. Great Pond. and Goodwin Pond Grab sampies of well Quarterly Two monitoring wells water Sediment Grab samples of stream Quarterly Five samples are taken along the bottom sediment last 137 m (50 ft) of the site creek; additional samples taken at Great Pond and at Goodwin Pond Soil Gran samples Semiannually Samples taken at 14 samphng stations around the piant site Vegetation Grao samoles Semiannuativ Samples taken at 18 locations around the site s
- Analyses were performed for uranium, gross alpha, gross beta. and r}amma 50ectra.
4-1
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Onsite tainpling locations.
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4-3 4.1.1.1 Air monitoring Each of the exhaust streams which may contain radionuclides vented into the atmosphere is monitored continuously using a sample system that meets the American National Standards Institute requirements as set forth in ANSI-N-13.1, 1969, Appendix A.
The sensitivity is such that detection of release levels are below the federal discharge limit of 4.0 x 10-12 uCi/ml (10 CFR Part 20).1 Other than routine air monitoring inside operating buildings, no other onsite air monitoring is performed.
The frequency of sampling and types of analyses are shown in Table 4.1.
4.1.1.2 Aqueous effluent monitoring Liquid wastes are collected in holding tanks at Buildings 5 and 17.
Before liquid wastes are released to the onsite creek, the tanks are sampled to ensure that the release levels do not exceed the discharge limit of 3 x 10-5 uCi/ml as set forth in 10 CFR Part 20.
Surface water Grab samples are collected quarterly from seven sampling points (Figs.
4.1 and 4.2 and Table 4.1).
Five sampling sites are located along the Farmington River, and two are ponds on the CE site.
Table 4.2 sunnarizes the analytical results from the nonradiological sampling program for 1975 through 1979.
River sediment samples are analyzed for radiological substances only.
Table 4.2. Summary of nonradiological efffuents from surface-water sampling stations shown in Fig. 4.1 Fluoride Nitrate
'com) epom)
Average 1979 6.4 0.16 3.9
(
1978 68 o.21 3.4 1977 7.o 0.15
- 8. 7 1976 7.3 0.13 2.1 1975 6.9 o.18 2.8 Range 1979
- 5. 3 - 7.1
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<*o-22.0 1978 6.3-73
<o.1 - 1. 05 1.1-17 o 1977 6.3 - 7. 7
<o.1 -o.3 7 1 4-95 o 1976 6,2 - 9.5
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OffSite Sampling locations.
4-5 Groundwater The two wells located on the site are sampled quarterly anu analyzed for pH, fluoride, nitrate, uranium, and alpha and beta radioactivity levels (Sect. 3.6.2).
These wells do not supply water to the fuel fabrication process nor are they used for drinking water purposes.
Sediment Five sediment samples are collected along approximately the last 17 m (50 ft) of the onsite creek before its confluence with the Farmington River. Siimples are also taken from the Great Pond and Goodwin Pond.
The sample frequency and types of analyses for the onsite aqueous samples are shown in Table 4.1.
4.1.l.3 Atmospheric fallout monitoring Precipitation and fallout are continuously collected at 14 sampling locations within the Windsor plant site boundaries.
Locations of sampling are shown in Fig. 4.1, and the frequencies of sampling and types of analyses are shown in Table 4.1.
4.1.1.4 Soil and vegetation monitoring Soil and vegetation are collected semiannually from 14 onsite locations (Fig. 4.1).
The types of sampling and analyses are described in Table 4.1.
4.1.2 Offsite monitoring program The offsite monitoring program conducted by the applicant is shown in Table 4.3.
The sampling locations are shown in Fig. 4.2.
4.1.2.1 Aouatic biota There are no monitoring programs for biota in either the CE onsite creek or in the Farmington River.
Accidental releases of acidic or alkaline process effluents would be intercepted by the industrial wastewater treatment facility and neutralized before release to the onsite creek.
These measures wculd minimize any potential impacts to aquatic biota in the onsite creek or Farmington River as the result of accidental releases.
I 4-6 Table 4.3. Offsite environmental monitoring program for the Combustion Engineering Windsor plant *
- Sample type Frequency Location Water monitoring Grab samples of surface Quarterly Four samphng points along water the Farmington River (one upstream and three downstream from onsite creek confluence)
Sediment Grab samples of stream Quarterly Four sampling pcints along bottom sediment the Farmington River (one upstream and two downstream from onsite creek confluence)
Soil Grao samples Quarterly Tobacco fields on north, south, east, and west of the site boundary Vegetation Grab samples Quarterty Same locations as soil samples
'No offsite air monstonng was performed.
' Analyses were performed for uranium gross alona. gross beta and gamma spectra.
4.1.2.2 Air monitoring l
No offsite air sampling is routinely performed as part of the environ-mental monitoring program, and the staff is of the opinion that no sampling is necessary.
4.1.2.3 Acueous release monitoring Quarterly grab samples are taken at four samoling points along the Farmington River. The samples are taken both upstream and downstream i
from the confluence of the onsite creek (the confluence is also sampled).
The sampling locations are shown in Fig. 4.2, and the types of sampling and analyses are described in Table 4.3.
l River sedime.at samples are collected at four locations on the Farmington River at one uostream and three downstream stations.
Samole locations are shown in Fig. 4.2, and the types of samoles and analyses are shown in Table a.3.
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f 4-7 4.1.2.4 Soil and vegetation monitoring Soil and vegetation samples are collected twice each year at four offsite Tocations.
The locations of the sampling sites are shown in Fig. 4.2, and the types of samples and analyses are shown in Table 4.3.
4.1.3 Mitigating measures Although the dose estimates for man resulting from the routine airborne and liquid releases of radionuclides to the environment are quite low and well below existing standards for safe operation (Sect. 4.2), it is important that an adequate program of environmental monitoring be main-tained to provide an early alert for potential problems and as an aid for keeping onsite and offsite exposures as low as reasonably achievable.
4.2 DIRECT EFFECTS AND THEIR SIGNIFICANCE 4.2.1 Air ouality Annual nonradiological emissions of various pollutants to the atmosphere are presented in Table 4.4.
Ground-level concentrations of the airborne pollutants were calculated assuming 240 working days (8 h/d) per year (ER, Appendix A, Fig. 3-5B) and using the x/Q value obtained by extrapo-lation of the values in Appendix A, Table A.4, southeast sector. This x/Q value represents a location at the site boundary southeast of the power plant, which is the major source of nonradiological emissions on the CE site. At this location, the site boundary most closely approaches the power plant (within about 440 m, ER Fig. 2.9), and the estimated off-site pollutant concentrations are maximal.
The power plant burns No. 6 fuel oil for heating and cooling all facilities on the site. The calcu-lated ground-level concentrations of pollutants (Table 4.4) are based on dispersion modeling that produces conservative results (Appendix A).
The calculated ground-level concentrations resulting from the CE facility alone are presented in Table 4.4 When the concentrations of S0x and N0x are added to their respective ambient concentrations found in the area (Table 3.1), the national air quality standards (Table 3.1) are not exceeded.
For particulates, the air quality region including the CE site is designated as a nonattainment area (Sect. 3.3).
However, the ambient levels of particulates at two monitoring stations in the area (Table 3.1) and emissions from the plant (Table 4.4) indicate that the standards for particulates will not be exceeded as a result of CE facility operations.
The concentrations of nydrocarbons and carbon monoxide are very low and would not cause violation of the air quality standards.
Several other pollutants listed in Table 4.4 are subject to control by the state of Connecticut.
Connecticut identifies acetone, isoprocyl alcohol, and Freon as photochemically unreactive and limits the discharge of each to 363 kg/d (800 lb/d). Assuming 240 working days in a year and
4-8 Table 4.4. Nontediological emimesono to the atmosphere
- Source Annual quantrty Annual average Effluent *
- n entra ns at 440 m dh 3
(t)
(yg/m )
Partculates' Heatmg and cooling' 4.6 5
SO, Hestmg and cooimg 45.4 41 Total NO, 22.1 20 NO, Heecng and coolmg 21.7 NO, Pickling operacon O.4 CO Hestog and cooling 2.9 3
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Total hydrocarbons 11.2 11 Hydrocarbons Heatmg and coolmg 0.6 1
Acetone Ceaneg operacons 1.6 2
Isopropyl alcohol Coeneng operations 3.3 4
Freon Ultrasonc cleenmg tank O.6 1
Perchlorethylene Degressmg tank 5.1 5
Heisum Weiding operations O.4 1
Argon Weidmg operatons 31.6 29 Nitrogen Nondestructive testmg 1.8 Hydrofluorm acid Pickling operanon Trace
'Estunated from factors for No. 6 fuel od from ref.10.
'Additeonel potonnel sources of poHutants include polyvmyl alcohol (2.7 t annuacy) and zmc stearate (0.6 t annuady) bumed off dunng dowaxmg operatons. Both of these are in the radio-logcal waste stream.
'In addition to radiologcal partculates, a smad amount of UO. Ups fines are towased to 2
the atmosphere.
dAll emissaons from heating and cooling represent consumpton of No. 6 fuel od for the entre site. Emisseons for Bidgs. 5 and 17 are less than one-half this amount.
Source: ER. Table 4-4; ER. Appendix a(7).
calculating from Table 4.4, discharges of each of these pollutants are less than 16 kg/d (35 lb/d). Connecticut identifies perchlorethylene as photochemically reactive and limits discharges to 18 kg/d (40 lb/d).
Calculated daily discharges are 21 kg/d (47 lb/d), which exceeds the Connecticut limit.
4.2.2 Land use Currently, there are no plans for construction or addition of major facilities at the CE Windsor site (ER, Sect. 2.1.5).
However, two land parcels totaling 223 ha (550 acres), most of which is agricultural, have recently been curcnased and added to the formerly 225-ha site.
There are no plans for industrial use of :nese lands, anc tne agricultural use will be allowed to continue in tne near future (ER, Sect. 2.1.5).
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4-9 Therefore, the proposed relicensing will result in no impacts on land use other than the continued use of the site for the existing industrial facilities.
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4.2.2.1 Mineral resources Continued operation of the plant facility will have no effect on the potential for or future extraction of any existing mineral resources at the plant site.
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4.2.3 Water i
4.2.3.1 Surface water Under normal operating conditions, the industrial wastewater discharge from Buildings 5 and 17 contains only hydrochloric acid, sodium hydroxide, and detergent solution (Table 4.5).
Because of the dilute nature and neutralization of these chemicals, the industrial wastewater is essen-tially unchanged in either physical or chemical quality from that of the
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municipal district water entering the site. As such, without treatment, it is discharged directly to the CE onsite creek via the CE industrial drain.
Table 4.6 contains a summary of 1980 discharge data which demon-strates compliance with the NPDES Permit Program.
4.2.3.2 Groundwater Continued operation of the plant facility will have no effect on the availability of potential groundwater resources and should not affoct groundwater quality.
l 4.2.4 Ecological i
4.2.4.1 Aauatic Because the water discharged from the CE site to the onsite creek is essentially unchanged from that entering the site from the municipal district water supply, there should be no effect on biota in the receiv-ing streams.
There are no nonradiological problems associated with the Farmington River as the result of discharge from the Windsor CE site l
(Mike Harder, Connecticut Water Compliance Unit, Hartford, Conn.,
personal communication to V. R. Tolbert, ORNL, Oct. 16,1981).
The only potential for a direct effect on aquatic biota lies in the accidental release of acidic or alkaline process water from the site.
Any effects on aquatic biota from such a release should be minimized by neutralization and dilution at the industrial wastewater treatment facility before discharge.
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4-11 Table 4.6. Summary of Cornbustion Engineering industrial drain discharge data for NPDES permit requirements Parameter monstored Samp4 Samp4 resWts AW troquency type range average Total suzpended sonds Weekly 24-h composite toppm 6 ppm Temperature Weekly 4-h average 50T-841 67.77 pH Weekly 24-h composite 6.o - 8.o
6.9 Source
ER 1981.
4.2.4.2 Terrestrial There are no plans for cnnstruction or addition of new facilities at the CE site (Sect. 4.2.2).
Also, no changes in land use are planned for the two new parcels of land that have been added to the site (Sect. 4.2.2).
Therefore, there will be no clearing of vegetation or loss of wildlife habitat associated with the proposed action.
Because endangered species do not frequent the site and vicinity (Sect. 3.8.1), none should be affected by continued operation.
Emissions to the atmosphere do not cause air quality standards to be exceeded (Sect. 4.2.1) and should have no significant impacts on vege-tation nor agricultural crops in the vicinity of the site.
4.2.5 Radiolooical imoacts The radiological impacts of the CE Windsor fuel fabrication facility were assessed by calculating the maximum dose to the individual living at the nearest residence and also at the nearest site boundary.
Except where specified, the term " dose" s referred to in this report is actually a 50-year dose commitment for all internal exposures; that is the total dose to the reference organ that will accrue from one year of intake of radionuclides during the remaining lifetime (50 years) of the individual.
It is conservatively assumed that the individual spends all of his time at the reference location and that all of the food consumed is produced on the site.
The dose reflects the release of radionuclides from the I
combined stack effluents. Where possible, site-specific data are used l
for calculating dose.
4.2.5.1 Doses from airborne releases l
Emissions from building exhaust stacks are monitored continuously, and the annual release from effluents from Buildings 5 and 17 for 1980 was 43.3 uCi.; On the basis of a 235U enrichment of 4.li,, the radionuclides released are shown in Table 4.7.
The nearest residence is aporoximately 700 m (0.44 mile) east of the site; the nearest site boundary is 560 m (0.35 mile) west of the release stacks.
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4-12 Table 4.7. Annual release rete of radionuctedes in the stock effluents of the Combustion Engeneering Windsor fuel febrication facility
- stual reisese p
C/yeM U-234 3.7oE-0 U-235 1.33E-6 U-238 5.coE-6
- Based on release rates after 1979 when the maxwnum ennenment was ricreased from 3.5 to 4.1% assU. (Sea USNRC Docket No. 70-11oOJ Doses were estimated using the computer code set forth in ORNL-5532.2 The methodology is designed to estimate the radionuclide concentrations in air; rates of deposition on ground surfaces; ground-surface concen-trations; intake rates via inhalation of air and ingestion of meat, milk, and vegetables; and radiation doses to man from the airborne releases of radionuclides.
With the code, the highest estimated dose to an individual in the area and to the population living within an 80-km (50-mile) radius of the site are calculated.
The basic equation used to estimate the dispersion of the airborne plume is the Gaussian plume equation by Pasquill3 as modified by Gifford."
Radionuclide concentrations in meat, milk, and vegetables consumed by man are estimated by coupling the output of the atmospheric transport models with the NRC Regulatory Guide 1.109, Terrestrial Food Chain Models.5 Parameters used in dose calculations are given in Appendix A.
Dose to the maximally exposed individual The 50-year dose commitment to the maximally exposed individual living at the nearest residences [700 m (0.44 mile) west of the plant site]
are shown in Table 4.8.
The total-body dose of 0.00053 millirem resulted primarily from the inhalation (91".) and ingestion (7.6%) pathways.
Approximately 86% of the dose was due to the m u and 10". to the 233U radionuclides released in the airborne effluents.
The highest organ dose, 0.013 millirem, was to the lungs.
The total-body and organ doses are only a smail fraction of those allowed by the apolicable NRC regulations, 500 millirems to *he total body, gonads, and bone, narrow and 1500 to the other organs (10 CFR Part 20).
4-13 Table 4.8. Fifty-yeer does commetment* to the maximally exposed individual at the neerest residence
- from the airborne effluents of the Combustion Engineering Windoor plant Dose (melbrem) p, Total body Bone Kidneys Lungs Inhalate'
- 4. 8E-4 8.2E-5 3.5E-4 1.3E-2 8
ingestion' 4.oE 5 4.oE-5 1.9E-5 2.oE-8 immersion-.n-air 9.8E-11 1.6E-Io 7.oE 11 8.5E-11 Exposure to surfaces 7.1E-6 1.2E-5 5.0E-6 5.9E-6 Total 5.3E-4 1.3E-4 3.7E-4 1.3E-2
' Fifty-year dose commitment from exposure to one year of plant operation.
' Nearest resadence is 700 m (o.44 mile) west of the site.
3
' Based on an inhalation rate of 8000 m / year.
Read as 4.8 X lo-*
d
' Based on maximum etake rates of 280 kg/ year of vegetables. 310 L/ year of milk, and 110 kg/ year of meat.
Similarly, the doses are well below the EPA standards, 25 millirems to the total body, 75 millirems to the thyroid, and 25 millirems to the other organs (40 CFR Part 190).
Additionally, the total-body dose to the maximally exposed individual from routine airborne releases. 0.00053 millirem, is only 0.00048% of the normal background radiation to area residents, 110 millfrems/ year.
Thus, the maximum doses to the nearest resident represent only a small increase in the radiation dose above background.
The doses at the nearest site boundary are shown in Table 4.9.
The doses are actually smaller than those received by the nearest residenc even though the site is closer to the point of release because the prevailing wind direction is toward the nearest residence.
f Dose to the cooulation within 80 km of the olant site The population dose commitments resulting from the routine annual releases of radionuclides (Table '.1) are shown in Table 4.10.
The total-body population dose was 0.00o6 man-rem, only 0.00023% of the population dose from normal radiation background of 3.66 x 10s man-rems.
An estimated 1981 pooulation of 3,504,062 persons live within the 80-km radius of the plant site (see Table A.8 in Appendix A).
The estimate is cased on the l
assumption that the area population grew at the same rate as the State of Connecticut (percentage cnange frem 1970 to 1980 was 2.5%).i
O 4-14 Table 4.9. Fifty-year dose commitment
- to the maximally exposed individual at the nearest boundary" from the airborne effluents of the Combustion Engineering Windsor plant Dose (maihrem)
Pathway Total body Bone Kidneys Lungs Inhalatm' 1.6E-4 2.7E-5 1.2E-4 4.5E-3 d
Ingestion' 1.4E 5 1.4E 5 6.9E-6 7.0E-9 Immersion-e-air 3.3E-11 5.5E 11 2.4E-11 2.9E-11 Exposure to surfaces 2.5E-6 4.3E-6 1.8E-6 2.1E-6 Total 1.8E-4 4.5 E-5 1.3E-4 4.5E-3
' Fifty-year dose commitment from the exposure to one year of plant operation.
' Nearest site boundary is 560 m (0.35 mile) east of the plant site.
3
' Based on an innalation rate of 8000 m / year.
- To be read as 1.6 X 10-d
' Based on maximum intake rates of 280 kg/ year of vegetables. 310 L/ year of milk, and 110 kg/ year of meat.
Table 4.10. Fifty-year dose commitment
- from airborne effluents to the population
- living within 80 km of the Combustion Engineering Windsor plant Population dose (manvem)
Pathway Total body Bone Kidneys Lungs d
innalation" 7.0E-3 1.2E-3 5.CE-3 1.9E-1 ingestion' 1.3E-3 1.3E-3
- 6. 7E-4
- 6. 7E-7 l
Immersion-n-air 1.4E-9 2.4E-9 1.0E-9 1.2E-3 Exposure to surfaces 2.5E-4 4.1E-4
- 1. 7 E-4 2.1E-4 Tota 6 8.6 E-3 2.9E-3 5.8E-3 1.3E-1
' Fifty-year dose commitment ' rom exposure to one year of f
plant coeration.
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'Sased on estimated 1981 occulation for the area of 3.504.062 persons.
3
' Based on an ennatation rate of 80CC m / year.
Read as 7.0 X 10~3
' Based on the average intane rates of 103 kgivear of vege-t tactes.110 L/vear of milk. and 95 kg/ year of meat.
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4-15 4.2.5.2 Doses from aqueous releases The methodology used for calculating the 50-year dose comitment: to man from the relaase of radionuclides to the aquatic environment is described in detail in ORNL-4992.7 Three exposure pathways are considered in dose determination: water ingestion, fish ingestion, and submersion in water (swimming).
Internal and external dose conversion factors are discussed in Appendix A.
The dietary intake rates are found TFNRC R5gulatory Guide l 1.109.s The release ratel and concentration in the Farmington River are shown in Table 4.11.
Table 4.11. Annual release rate of radionuclides in the liquid effluents from the Combustion Engineering Fuel Fabrication Facility
- and the concentration in the Farmington River
- Release rate Concentraten in (gCi/ year) nver (uCi/cc) 23'U 1.36E3' 1.4E-12 23sU 4.88E 1 5.oE-14 23'u 1.84E2 1.9E-13
' Based on release rate after 1979 wnen the maximum ennch-ment was increased from 3.5 to 4.1% U-235. (See USNRC cocket No. 70-11oo)
Average annual flow of the Farmington River is 9.71 X to
8 cc! year.
3
' Read is 1.36 X 10.
Dose to the maximally exoosed individual The 50-year dose commitments for the individuals exposed to varicus aquatic pathways associated with the use of the Farmington River water are shown in Table 4.12.
Of the total-body dose 0.0026 millirem, approximately 78". is due to the ingestion of water with essentially the remainder due to the ingestion of fish.
All organ doses are well below the limits established in 10 CFR Part 20, 500 millirems / year to the total body, gonads, and bone marrow, and 1500 millirems / year to the otner organs.
Similarly the doses are only a small fraction of the EFA standards, 25 millirems to the total bcdy, 75 millirems to the thyroid, and 25 millirems to tne otner organs (20 CFR Part 190). Additionally,
4-16 Table 4.12. Maximum 50-year does commetment' to individuals from the routine release of liquid effluents into the Fermington River Aquatic Pathways Total body Bone
- Kidneys Lungs Submersion ai water' 7.7E-Iod 1.oE-9 6.7E-lo 7.3E-1o Ingestion at water
- 2.oE-3 2.1E-3 1.CE-3 1.oE-6 Ingesticn of fish' 5.8E-4 6.1E-4 2.oE-4 2.9E-7 Total 2.6E-3 2.7E-3 1.3E-3 1.3E-6
' Internal doses are So-year dose commitments for one year of radionuclide intake.
- Endastead cetis of the bone.
' Based on swimming in the nver for 1% of the year.
Read as 7.7 X lo~'O
' Based on water intake (maxrnum value) of 730 L/ year.
' Based on fish consumption (maxwnum value) of 21.o kg/ year.
the total-body dose, 0.0026 millirem, is only 0.0024% of the natural background dose (110 millirems per year) in the vicinity of the plant.
Population dose commitments from liouid effluents The Farmington River is not known to be used for human consumption below the plant from the onsite creek outfall to its confluence with the Connecticut River.
Because none of the municipalities in the area use the river as a source of drinking water, the radiological impacts on the regional population is considered minimal.
4.3 INDIRECT EFFECTS AND THEIR SIGNIFICANCE i
4.3.1 Socioeconomic effects The request for license renewal by CE involves neither expansion nor l
curtailment of the existing fuel fabricating facilities.
Therefore, the I
size of the current number of operational employees should not change significantly, thus eliminating the potential for significant socio-
[
economic impacts in the surrounding communities.
I i
4-17 4.3.2 Potential effects of accidents 4.3.2.1 Accidents involving radioactive materials Accidental criticality The staff considers an accidental criticality event extremely unlikely because of the large quantities and discrete geometries required to achieve criticality with low-enriched material. Uranium fuel fabrica-tion plants including the CE facility are required to be designed to minimize the potential for a criticality accident, and such an accident has not occurred in currently operating facilities.
Notwithstanding the above, the consequences of such an accident are evaluated here assuming an initial burst of.10 8 fissions, which is 1
equivalent to the release of approximately 32 MW. This is a much larger excursion than could be expected in any system in the fuel fabrication facility since no solutions containing enriched material are used in the facility. The initial event of 1018 fissions releases 5.71 x 106 calories of heat.
If dry UO2 were involved (a minimum of approximately 225 kg in discrete geometry), a temperature rise of approximately 500 C would occur and cause powder to disperse because of air expanding in the pore 2 (pellets) would expand but would not be otherwise volume.
Solid UO a ffected. NRC Regulatory Guide 3.348 postulates this event followed at 10-min intervals by additional bursts of 1017 fissions for a total of 1019 fissions.
The potential radiation dose at the nearest residence, 700 m (0.44 mile) from the orocessing building, was calculated using NRC methodology.a No credit was taken for structural attenuation.
To assess the maximum potential dose, the folicwing assumptions were made:
Pasquill type F atmospheric stability, transport time to nearest residence with the cloud moving at a rate of 1 m/s, and no deviation of centerline from the plume of the cloud. The doses from prompt gamma and neutron releases were also calculated. The results are given in Table 4.13.
Table 4.13. Calculated maximum dose potential at the nearest residence from airborne radionuclides resulting from a criticality event of lo" fissions l
Nearest res dent dose frem)
Whole bmsy 2.38 Surface A33 I
Thyroid 5.71 l
4-18 The effective delay time is based on two factors: decay time in the fuel fabrication building during release and transit time to the nearest residence. The decay half-life for Building 17 is 5.2 min, assuming criticality occurs in the pellet shop, where the ventilation rate is 1100 m3/ min (38,862 ft / min) and the building volume is 8239 m3 3
(291,000 ft ).
3 The delay after release is 11.7 min, which is based on the cloud moving at 1 m/s for 300 m.
No measurable effects would be observed from these calculated doses in Table 4.13.
It is probable t5t the true dose would be much lower in an actual accident.
Fire in the HEPA filters The applicant uses HEPA filters fabricated from metal frames and non-combustible filter media and, therefore, has eliminated potential fire occurrence in the HEPA filter system.
Utility failures Loss of externally supplied electric power and water supply are considered credible accidents.
The applicant maintains an onsite diesel-powered emergency generator for sintering furnaces, HEPA filtration exhaust fans, and other critical equipment to minimize loss of power impact.
The applicant states " power outages rarely occur." The staff notes that if the standby generator were also to fail, the only effect would likely be the damage of hot equipment by warping. Because there would be no forces to disperse uranium to the surroundings, a complete power failure would not affect the environment.
As a result of electric power failure or of shutdown for maintenance, failure of the heating, ventilating, and air conditioning (HVAC) system is postulated as a credible accident.
This type of accident could cause some elevation of radioactivity levels within the plant, thus requiring close surveillance by health physics personnel, perhaps evacuation of certain plant areas, and/or the use of respiratory protection apparatus.
However, such conditions should not produce any hazard to the environment or to inhabitants outside the plant; with the HVAC system shutdown, no radioactive effluents would be exhausted from the facility.
Water will be supplied througn a 30.5-cm (12-in.) main by the Metropolitan District Commission.
If a water main break occurs and causes the total plant water supply to cease, sufficient water is available through storage in the cooling systems to operate the plant long enougn to provice time for an orderly shutcown.
4-19 In the event of a complete failure of the fire protection water supply line, a 1609-m3 (425,000-gal) storage tank is maintained at all times to provide water for two separate tanks for fire-fighting purposes.
Radiological or chemical releases to the environment are not expected to increase above regulating limits as a result of electric power failure, HVAC shutdown or failure, or water loss.
Hydrogen explosion in sintering furnace Some minor furnace explosions have occurred within existing fuel fabri<
cation plants.
The most prevalent cause of events of this type has been hydrogen " pops" (i.e., hydrogen and air reactions), which have resulted from insufficient purging of residual air pockets during startups and shutdowns. The applicant is currently using cracked amonia instead of hydrogen gas in the sintering furnaces, which greatly reduces a potential explosion. An explosion of the magnitude postulated below is considered incredible.
For purposes of this analysis, it is hypnthesized that precautionary steps fail and that an explosion occurs in one of the furnaces. The resultant force would not be sufficient to destroy the furnace, but uranium could be blown out. The furnace could contain up to 280 kg of uranium in the form of UO2 pellets.
It is assumed that all of the uranium is blown out of the furnace and that 1% (ref. 9), or 2.80 kg, of UO is reduced to powder and is drawn into the building exhaust system.
2 With a filter efficiency of 99.9%, the calculated atmospheric release to the environment is 2.80 g of uranium (8.5,:Ci).
The calculated potential dcse to an individual at the nearest residence is 2.1 x 10-3 mSv (0.21 millirem) to the lungs and 1.8 x 10-5 mSv (7.8 x 10-3 millirems) to the bone.
4.3.2.2 Accidents involving nonradioactive materials Chemicals are stored in bulk quantities at the site.
Spills, fires, and I
explosions are the main accident mechanisms through which stored chemicais may impact the environment.
These accident scenarios, their environmental effects, and plant design features to minimize such effects are discussed
- below, l
Major soills The chemicals potentially subject to spills include nitric acid, anhydrout ammonis, liquid nitrogen, liquid procane, and spent acid. The acclicant has employed the following measures to ensure tne integrity of chemical storage tanks:
proper design and operation, safety relief valves, i
remote location, and routine visual inscection by trained plant perscnnel.
l The apolicant plans to remove the 19,000-L (5000-gal) stainless steel
4-20 bu% tank from the plant and begin using ll4-L (30-gal) commercial
..;inless steel drums.1 Plans are being made by the applicant to design dnd construct a containment dike around the spent acid storage tanks.
The chance of a major spill of the contents of a tank is considered remote. Should such a spill occur, appropriate measures will be taken to recover or neutralize the spilled material. Therefore, accidental spills are not expected to impact offsite waters.
Fires and explosions Potential sources of fires or explosions are acetone, alcohol, zirconium
" fines," and liquid propane.
Nonflammable gases such as nitrogen, argon, and helium will also be stored as cryogenic liquids. Acetone and alcohol are contained in 3.785-L (1-gal) safety cans equipped with flame arrestors inside the facility.
Solvents are limited to grounded drums located outside the buildings.
Zirconium " fines" are a potential source for fires. As a precaution, zirconium is limited to 2.3 kg (5 lb) at machining locations and is submerged in water.
Propane is highly flam-mable and would present a fire hazard should a large leak occur. The fire, however, would be restricted to the immediate vicinity of the tank; no significant environmental effect would result.
Accidents of this type have a very low probability of occurring but may result in the release of some radioactivity to the immediate vicinity of the plant environs.
The probability of a fire with release of radioactivity from Buildings 5 or 17 has been minimized through carefully engineered safeguards, strict control of combustible materials, and protective measures to control a fire if it does occur. Safeguards include extensive sprinkler systems, a 1.61 x 106-L (425,000 gal) storage tank with backup pumping capability, i
external fire hydrants, and portable fire extinguishers within the l
buildings.
l For purposes of this analysis, it is hypothesized that precautior.ary steps fail and a fire occurs in the fuel storage area.
The staff esti-mates a maximum loss of uranium (4.1", 235U) of 60 mg per fuel rod.1C The activity released is 0.18 uCi per fuel rod.
Assuming a fire does
(
occur, it is estimated that all 300 fuel rods could rupture, which is i
tne maximum number of fuel rods in a storage container.
This could result in the release of 55 uCi.
It is conservatively assumed the total 55 uCi of uranium is released frcm the building with failure of the fil :ers.
The nearest resident [700 m (0.44 mile) away] from Building 17 v-1d receive a calculated dose equivalent of 1.36 x 10-2 mSv (1.36 sms) to the lungs and 1.16 x 10 " mSv (1.16 x 10-2 millirems) to ti
(
These doses are only 0.27 and 0.002", of tne 10 CR Part 20 stai.
for lungs anc bone, respectively.
l i
4-21 4.3.2.3 Transportation accidents Shipments of nonradioactive materials All hazardous and flammable materials will be packaged and shipped in accordance with U.S. Department of Transportation regulations and any applicable state and local laws.
Even so, collisions and other acci-dents that could threaten human life may occur.
Published accident statistics yield an accident probability between 1 x 10-6 per kilometer (1.6 x 10-6 per mile) and 1.6 x 10-6 per kilometer (2.6 x 10-6 per mile)
(ref.11) with risks to human life being 0.51 injuries per accident and 0.03 fatalities per accident.
The staff has estimated about 150 shipments per year. On this basis and the estimated total shipping distance involved [24,000 km/ year (15,000 miles / year)], the probability of an accident involving shipments of nonradioactive materials is between 0.024 and 0.039 per year. The estimated injury and fatality rates associated with this accident rate are from 0.012 to 0.020 injuries per year and 0.0007 to 0.039 fatalities per year.
The nonradioactive material shipments to the applicant's plant do not present any unique transportation hazard over and above those handled by nonnuclear manufacturing facilities utilizing similar chemical reagents.
i Shioments of radioactive materials In addition to the collision and other accident hazards addressed above, transportation accidents may present radiological hazards to the public.
A significant release of uranium from a fuel assembly shipping accident is incredible because of the physical form of the uranium (pellets) and because of the protection offered by the enclosed tubing, the fuel assembly structure,ll and the shipping container. A further mitigatory factor is the very low specific activity of the unirradiated fuel.
Uranium dioxide powder has a very high density, which would make dispersion difficult.
The transport of enriched uranium is strictly regulated by the U.S.
l Department of Transportation,12-15 and package design and specifications must be approved by NRC.
Containers must be designed to withstand l
specified hypothetical accident conditions acolied secuentially in an l
order specified in the regulations to determine the cumulative effect on the container being tested.
Criteria include free drops, punctures, thermal stress, and water immersion tests. These tests, which are more severe than any expected transportation accidents, make the probability of release of contents or accidental criticality incredible.
In addi-tion, the applicant must establish, maintain, and execute a quality 12 assurance prcgram tnat satisfies applicable criterial: to ensure that all packages are procerly pre::arec for snicment.
l
4-22 The environmental effects of transportation accidents involving prope.y packaged radioactive materials have been thoroughly analyzed and documented.9'11:16 17 These analyses show that the radiological risk from transportation accidents involving radioactive materials does not contribute appreciably to the accident consequences (less than 1/2 of 1 ". ).
The estimated annual shipments of radioactive materials to and from the Windsor site are 45 and 35 shipments, respectively).
The average shipping distance has been estimated at 1609 km (1000 miles. The combined f
shipping distance of the 80 radioactive material shipments to and from the site is therefore estimated to be 128,720 km/ year (80,000 miles / year).
Using the accident [between 1 x 10-6 per kilometer (1.6 x 10-6 per mile) and 1.6 x 10-6 per kilometer (2.6 x 10-6 per mile)], injury (0.51 per accident), and fatality (0.03 per accident) frequencies noted above, the probabilities of collision-related injury and fatality for radio-active material shipments would range from 0.066 to 0.105 injuries and 0.0039 to 0.0062 fatalities per year and average 0.086 injuries and 0.005 fatalities per year.
4.3.2.4 Natural phenomena Tornado The probability of tornado occurrence has been estimated as once in a million years (Sect. 3.2.3).
Were the CE facility to be struck by a tornado of sufficient force to damage a uranium-containing building, radioactive material might be released to the atmosphere.
The ultimate fate of this material is impossible to predict with accuracy. However, conservative estimates of the ground-level airborne concentrations of UO: resulting from a tornado striking the proposed facilities were made.
Maximum ground-level UO: concentrations are predicted to be 0.15 mg/m3 and are predicted to occur 500 m from the facility boundary.
Such concentrations could occur for approximately 200 s before decreasing to less than 5% of the predicted maximum value. Such a concentration of U0- would lead to a maximum decosition of 0.02 g/m2 and affect an area 2
of less than 0.15 km. The staff estimates the 50-year dose commitment from inhalation to be less than 50 mSv (0.5 millirem) to the total body, 15 mSv (0.15 millirem) to the lungs, and 10 mSv (0.1 millirem) to the bone from such an event.
The estimated maximum ground deposition, 2
0.06 aci/m, might recuire remedial action.
Earthcuake The chance of a major earthquake occurring near the site is imorobable.
The CE Windsor clant site is located in a cne i eartncuake intensity, corresconding to Intensity '/I on the modified 'dercalli scale of 1931.
The CE plant was structurally designec to withstana Zone 2 earthquake intensities, wnicn are more severe tnan nose of Ione 1.
4-23 Windstorm and flood The possibility of the occurrence of a severe windstorm due to a hurricane in the immediate area of the site that could result in major property damage or radioactive material release is negligible. The nuclear fuel fabrication plant and nuclear laboratories are designed to withstand steady winds of about 161 km/h (100 mph), whereas the highest recorded wind velocity was 113 bn/h (70 mph) in November 1950. Hurricane winds would not cause appreciable damage to the plant because it is located far enough inland.
The elevation of the facility is 55 m (180 ft) above mean sea level (msl);
the highest recorded flood in the area, which occurred in August 1955, was 34 m (110 ft) msl. Therefore, the likelihood of any damage associated with floods or floodwater of tributaries is highly improbable.
4.3.3 Possible conflicts between the crocosed action and the objectives of federal, regional, state, and 1ccal plans and policies At this time, the staff is not aware of any conflict between the proposed action and the objectives of federal, regional, state (Connecticut), or local (Windsor) plans, policies, or controls for the action proposed as long as proper agencies are contacted, proper applications are submitted (ER, Sect. 9), and proper monitoring and mitigatory measures are taken to protect the environment and public health and safety.
4.3.4 Effects on urban cuality, historical and cultural resources, and society The environmental effects of the proposed license renewal action as dis-cussed above are considered to be insignificant. There may be adverse effects on urban quality if reactor fuel were not available.
The facility has not affected historical or cultural resources.
The short-term societal effects during operation are and will be minimal, t
and there will be minimal effects after decommissioning and reclamation because the site then will be required to meet federal standards for l
unrestricted use.
l i
I
4-24 REFERENCES FOR SECTION 4 1.
U.S. Nuclear Regulatory Commission, Environmenect Impace Infomation for the Combustion Engineering, Inc., Windsor, Connecticut, ?lant Site, Nuclear Fuel ?chrication Facility Nuclear Laboratories, Docket No. 70-1100, April 1981.
2.
R. E. Moore et al., AIRDCS-EPA: A Computerized Methodology for Eatinating Enviromental Concentrations and Dose to Man frcm Air-borne Releases of Radionuclides, ORNL-5532, Oak Ridge National Laboratory, Oak Ridge, Tenn., June 1979.
3.
F. Pasquill, "The Estimation of the Dispersion of Windborne Material,"
Meteoral. Mag. 90: 33(1961).
4.
F. A. Gifford, Jr., "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," Nucl. Saf. 2(4): 45-57 (1961).
5.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, Calculation of Annual Doses :o Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Ccm:liance with 10 CFR Part 50, Appendix I, Rev. 1, Office of Standards Development, Washington, D.C.,
1977.
6.
U.S. Department of Commerce, "1980 Census Counts of the Population of States, by Region and Division," Commerce Jees, Bureau of the Census (1981).
7.
G. G. Killough and L. R. McKay, A Methodology for Calcula ing Radiation Doses frcm Radioactivi:y Released :o che Environment, ORNL-4992, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 1976.
8.
U.S. Nuclear Regulatory Commission, Office of Standards Develcoment,
" Assumptions Used for Evaluating the Potential Radiological Conse-quences of Accidental Nuclear Criticality in a Uranium Fuel Fabri-cation Plant," '.S. Nuclear Reguia:ory Ccmmission Recula:ory Guide, J
Regulatory Guide 3.34 Rev. 1, Washington, D.C., July 1979.
l 9.
U.S. Atomic Energy Commission, Directorate of Licensing, Environ-menza! Eurvey of the Uraniwn Fus; cyc;e, Report WASH-1248, Sect. E, i
Washington, D.C., April 1974.
- 10. Enviremen:c; Re:cr:, Alabama Nuclear Fuel Fabrication Plant, p. 5-4, Docket No. 70-29'9, Decemcer 1979.
0 11.
U.S. Atcmic Energy Commission, Directorate of Regulatory Standards, Envircmen a: Eurvey of :ransycr:::icn cf Radiocc ive Ma:eric:a :o and frcn Juclear Feuer ?! ants, Report WASH-1238, Cecember 1972.
l l
l
4-25 12.
13.
- 14. 49 CFR Parts 170-79.
15.
U.S. Nuclear Regulatory Comission, Regulatory and Other Responsi-bilities as Related to Transportation Acciden:s, Report NUREG-0179, Washington, D.C., June 1977.
16.
U.S. Nuclear Regulatory Comission, Fina! Invironmental Sea:ements on :he Transportation of Radioactive Ma:em*al by Air and Other Modes, Repor't NUREG-0170, Vols. 1 and 2, Washington, D.C., December 1977.
17.
U.S. Nuclear Regulatory Commission, Environmental Survey of Trans-porea: ion of Radioac:ive Materials to and from.Vuclear Pouer Plants, Report NUREG-75038, Suppl.1, Washington, D.C., April 1975.
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L
Appendix A METHODOLOGY AND ASSUMPTIONS FOR CALCULATING RADIATION DOSE COMMITMENTS FROM THE RELEASE OF RADIONUCLIDES
Appendix A METHODOLOGY AND ASSUMPTIONS FOR CALCULATING RADIATION DOSE C0ffiITMENTS FROM THE RELEASE OF RADIONUCLIDES A.1 METHODOLOGY AND ASSUMPTIONS FOR AIRBORNE RELEASES A.l.1 Methodology The radiation dose commitments resulting from the atmospheric releases of radionuclides are calculated using the computer code in ORNL-5532.1 The methodology is designed to estimate the radionuclide concentrations in air; rates of deposition on ground surfaces; ground-surface concentra-tions; intake rates via inhalation and ingestion of air and ingestion of meat, milk, and fresh vegetables; and radiation doses to man from the airborne releases of radionuclides.
Witn the code, the highest estimated dose to an individual in the area and the doses to the population living within an 80-km (50-mile) radius of the plant site are calculated. The doses may be sumarized by radionuclide, exposure mode, or significant organ of the body.
Many of the basic incremental parameters used in ORNL-55321 are conserva-tive; that is, values are chosen to maximize intake by man. Many factors that would reduce the radiation dose (e.g., shielding provided by dwell-ings and time spent away from the reference location) are not considered.
It is assumed that an individual lives outdoors at the reference location 100% of the time.
Moreover, in estimating the dose to an individual via ingestion of vegetables, beef, and milk, all of the food consumed by the individual is assumed to be produced at the reference location specified in the calculation. Thus, the dose estimates calculated by these methods are likely to be higher than the doses that would actually occur.
l The basic equation used to estimate the discersion of an airborne plume l
is the Gaussian plume equation of Pasquill2 as modified by Gifford.3 Radienuclide concentrations in meat, milk, and vegetables consumed by humans are estimated by coupling the output of the Atmospheric Transport Models with NRC Regulatory Guide 1.109, Terrestrial Food Chain Models."
The models are described in ORNL/TM-6100.5 A.l.2 Radiation exoosure oat uays and dose conversion factors Environmental transport links the source of release to the receptor by numerous exposure pathways.
Figure A.1 is a diagram of the most important pathways that result in the exposure of humans to radioactivity released to tne environment. The resulting radiation exposures may be either external or internal.
External exposures occur when the radiation source is outsice the irradiated bcdy, and internal exoosures are tnose frcm r3cicactive materials within the irradiated bocy.
A-3 ll.
A-4 ES 377 OIRECT ATMO5m*ERic ACUATIC
~
WIRAOIATION RELEASES RELEASES
! /
d o o
IMMERSION b
28 R5m Co AM T o
EXTERNAL uAu l
ATMCSPMERIC AQUATIC RELEASES NLEASES
- r T
E Soit y 3 5 4 j
z t
f 4r f
s TERRESTRIAL POTABLE FISH AND
~
N YEGETATION WATER SEAFCCOS INHALATION i
/
%,E Cef g 3 5 o"
ff
- r 2 2 d
l ec
\\
I WAN INTERNAL Fig. A.l.
Pathways for exDosure of humans from releases of racicaC:ive effluents.
A-5 The dose conversion factors for converting the radiation exposures to estimates of dose are calculated using the latest dosimetric criteria of the International Comission on Radiological Protection (ICRP) and other recognized authorities.
A.l.2.1 External dose conversion factors Releases of radioactive gases and particulates to the atmosphere may result in external dose by exposure to and/or immersion in the plume and in contamination of land surfaces. The dose conversion factors have been computed as summarized by Kocher in ORNL/NUREG-79, and those used in this report are shown in Table A.l.
Table A.1. Dose conversion factors
- for external exposure pathways Organ Radionuchde Total body Bone Kicneys Lungs Expoture to ground surface 2
(millirem / year per gCl/cm,
23'U 7 o7E2 2.95E2 1.oCE2 1.74E2 235U 1.53E5 2.o7E5 1.31E 5 1.39E5
- ssu 5.66E2 2.09E2 5.8E 1 1.21E2 Immersion in air 3
(mellirem/ year per gCi/cm )
23'U 6.77E5
- 7. loES 3.74E5 4.11ES 235U 6.84E8 9.36ES 5.92E8 6.33E8 l
23au 4 59E5 4.51E5 2.19E5 2.SoES l
Submersion in water 3
(millirem / year per aCi/cm )
23'U 1.65E3 1.69E3 8.38E2 9.81E2 235U 1.53E6 2.09E6 1.32E6 1.41E6
- ssU 1.13E3
- 1. toe 3 5.29E2 6.11 E2
'From CRNL NUREG-79 tsee Ref. 6).
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A-6 A.l.2.2 Internal dose conversion factors Factors for converting internal radiation exposure to estimates of dose have been computed and surmiarized by Dunning et al. in ORNL/NUREG/TM-190/V27 implementing recent models.8,9 The dose conversion factors in this report are presented in Table A.2.
These factors are data input into a computer l
code, which is used to calculate the dose from inhaled and ingested radionuclides.
Table A.2. Dose conversion factors
- for internal empow.re pathways Organ Radionuchoe Total body Bone' Kidneys Lungs Inhalation' (rem /gCa 23'U 1.70E1 2.90E0 1.40E0 4.72E2 23su 1.50E 1 2.SOE0 1.30E0 4.29E2 23sU 1.50E 1 2.40E0 1.30E0 4.29E2 Ingestion (rem /gCa 234U 1.70E0 1.8cEO 5.80E-1 8.23E-4 23su 1.60E0 1.50E0 5.30E-1 1.37E-4 23eu 1.50E0 1.50E0 7.60E-1 8.04E-4
'From ORNL NUREGiTM-190/V2 (see Ref. 7).
- Endosteal ceils.
- A5sumed a particle Size of 0.3 gm and solubihty class Y.
A.l.3 Radiation dose to the individual Internal exposure continues as long as radioactive material remains in the body, whicn may be longer than the duration of the individual's residence in the contaminated environment.
The best estimates of the internal dose resulting from an intake are obtained by integrating over the remaining lifetime of the exposed individual; such estimates are called " dose commitments." The remaining lifetime is assumed to be 50 years for an adult.
External doses are assumed to be annual doses.
The cose rates above the contaminated land surface are estimated for a height of 100 cm.
Following the initial deposition of radionuclides, the potential for exposure of humans may persist, decending on the influence of environ-mental redistribution, long after the clume leaves the area.
Concentra-tions of radionuclides at the coint of aeposition normally are reduced
A-7 by infiltration of radionuclides into the soil, by loss of soil particles through erosion, and by transport in surface water and in groundwater.
When the effects of these processes cannot be quantified, a conservative estimate of dose due to external exposure to contaminated surface is obtained by assuming that the radionuclide concentrations are diminished by radioactive decay only.
The dose is estimated for individuals at the nearest site boundary at the nearest residence. The intake parameters used for individual dose determination are shown in Table A.3. -
Table A.3. Intake parameters (adult)* used in lieu of site-specific data Maximum exposed Average exposed Pathway Mi l
ind:M ual*
Vegetables (kgiveart 281*
19o Milk (L/ year) 31o llo Meat (kg/yer) 11o 95 Orinking water lL/ year) 73o 370 Fisn (kg/ year) 21 6.9 3
inhalation (m / year) 8000 800o
'From Regulatory Guide 1.109.
'Used for calculating population doses.
' Vegetables only Uncluded leafy vegetables).
A.l.4 Radiation dose to the oopulation l
The total dose received by the exposed population is estimated by the summation of individual dose estimates within the population.
The area within the 80-km (50-mile) radius of the site is divided into 16 sectors (22.5 each) and into a number of annuli.
The average dose for an individual in each division is estimated, that estimate is multiplied by the number of persons in the division, and the resulting products are summed across the entire area.
The unit used to express the population dose is man-rem.
For this report, the population dose estimates are calculated for a population composed entirely of adults.
The parameters i
used for calculating population doses are shown in Table A.3.
A.2 METH000 LOGY AND ASSUMPTIONS FOR ACUEOUS RELEASES The metnodology used for calculating the 50-year dose conmitments to humans from the release of radionuclices to an aquatic environment is described in detail in ORNL J992. 3 Reference 10 also gives bicaccumula-tion factors for racianuclides in fresh-water fisn and sample proolems.
l
A-8 AQUAMAN is a computer codell that can also be used for calculating similar dose commitments from exposures to aquatic pathways.
Three aquatic exposure pathways are considered in dose determination:
water ingestion, fish ingestion, and submersion in water (swiming).
The internal dose conversion factors for converting exposure to dose are discussed in Section A.l.2, and the factors are shown in Table A.2.
The external dose conversion factors are shown in Table A.l.
Intake param eters. are shown in Table A.3.
A.3 ATMOSPHERIC DISPERSION The atmospheric dispersion model used in estimating the atmospheric transport to the terrestrial environment is discussed in detail in ORNL-5532.1 For particulate release, the meteorological x/Q values are used in conjunction with dry deposition velocities and scavenging coefficients to estimate air concentrations and steady-state ground concentrations. The atmospheric dispersion model estimates the concen-tration of radionuclides in air at ground surfaces as a function of distance and direction from the point of release.
Site-specific average annual meteorological data are supplied as input for the model.
Radio-active decay during the plume travel is taken into account in the computer code.1 Daughters produced during plume travel are calculated and added to the source term.
The area surrounding the plant site is divided into 16 sectors by compass direction.
Each sector is bounded by radial distances of 0.56, 0.7, 1.2, 2.4, 4.0, 5.6, 7.2,12.0, 24.0, 40.0, 56.0, and 72.0 km from the point of release.
Each distance represents the midpoint of a sector and x/Q values are calculated for each sector.
The x/Q values are shown in Table A.4.
Concentrations in the air for each sector are used to calculate dose via inhalation and submersion in air. The ground deposits result in external gama dose and, in addition, are assimilated into food and contribute dose upon ingestion via the food chain.
The meteorological data recuired for the calculations are joint frequency distributions of wind velocity and direction sumarized by stability class. Meteorological data from the nearby Bradley International Airport (Tables A.5 and A.6) are used to calculate the concentrations of radionuclides at a reference point per unit of source strength.
Depletion of the airborne plume as it is blown downwind is accounted for in the computer code by taking into account the deposition on surfaces by dry deposition, scavenging, and radioactive decay.1 Other parameters used in determining air concentrations are shcwn in Table A.7.
The copulation distribution by distance and sector usec to calculate doses given in Tacle 4.10 is shown in Table A.S.
o A-9 Table A.4. Ground-level CHl/Q values at vermus distences in each compass direction Distance CHl/O toward indscated direction 3
(m)
(s/m )
N NNW NW WNW W
WSW SW SSW 560 0.604E 5 0.291E-5 0.173E-5 0.968E-6 0.876E-6 0.979E-6 0.172E-5 0.278E-5 700 0.436E-5 0.213E-5 0.129E-5 0.720E-6 0.656E-6 0.737E-6 0.130E-5 0.208E-5 1.207 0.184E-5 0.915E-6 0.564E-6 0.314E-6 0.288E-6 0.326E-6 0.575E-6 0.914E-6 2.414 0.603E-6 0.303E-6 0.187E-6 0.103E-6 0.951E-7 0.108E-6 0.191 E-6 0.303E-6 4.023 0.273E-6 0.137E-6 0.847E-7 0.464E 7 0.426E-7 0.486E-7 0.863E-7 0.137E-6 5.633 0.165E-6 0.833E-7 0.511 E-7 0.278E-7 0.255E-7 0.292E-7 0.519E-7 0.821E-7 7.242 0.113E-6 0.569E-7 0.349E-7 0.189E-7 0.173E-7 0.198E-7 0.352E-7 0.557E-7 1.2070 0.528E 7 0.267E-7 0.162E-7 0.865E-8 0.788E-8 0.913E-8 0.162E-7 0.256E-7 2.4140 0.191E-7 0.964E-8 0.571E-8 0.293E-8 0.264E 8 0.314E-8 0.554E-8 0.875E-8 4.0234 0.862E-8 0.431E-8 0.247E-8 0.122E-8 0.109E-8 0.133E-8 0.233E-8 0.370E-8 5.6327 0.478E.8 0.235E-8 0.129E-8 0.621E-9 0.545E-9 0.679E-9 0.119E-8 0.190E-8 7.2420 0.283E-8 0.135E-8 0.695E-9 0.326E-9 0.281E-9 0.354E-9 0.619E-9 0.101E-8 S
ENE NE NNE 560 0.740E-5 0.470E-5 0.426E-5 0.395E-5 0.339E-5 0.244E-5 0.144E-5 0.179E-5 700 0.548E-5 0.356E-5 0.320E-5 0.301E-5 0.262E-5 0.186E-5 0.104E-5 0.130E-5 1.207 0.239E-5 0.159E-5 0.142E-5 0.136E-5 0.119E-5 0.842E-6 0.445E-6 0.556E-6 2.414
- 0. 793E-6 0.537E-6 0.476E-6 0.460E-6 0.405E-6 0.285E-6 0.146E-6 0.183E-6 4,023 0.360E-6 0.246E-6 0.218E-6 0.211E-6 0.186E-6 0.131 E-6 0.665E-7 0.828E-7 5.633 0.218E-6 0.150E-6 0.133E-6 0.129E 6 0.114E-6 0.799E-7 0.403E-7 0.501 E-7 7.242 0.149E-6 0.103E-6 0.909E-7 0.879E-7 0.777E-7 0.548E-7 0.276E-7 0.343E-7 1.2070 0.693E-7 0.485 E-7 0.430E-7 0.416E-7 0.368E-7 0.260E-7 0.131E-7 0.161E-7 2.4140 0.245E-7 0.176E-7 0.157E-7 0.151E-7 0.134E-7 0.955E-8 0.487E-8 0.587E-8 4.0234 0.108E-7
- 0. 787E-8 0.709E-8 0.673E-8 0.595E-8 0.430E-8 0.225E-8 0.266E-8 5.6327 0.576E-8 0.42SE-8 0.387E-8 0.363E-8 0.318E 8 0.233E-8 0.126E-8 0.117E-8 7.2420 0.321E 8 0.237E-8 0.221 E-8 0.201 E-8 0.173E-8 0.130E 8 0.753E-9 0.8 2E-9 l
[
t A-10 Table A.5. Frequency of a:---
'1-ic stability cassees for each direction Sector Fracton of tme in each stability class A
B C
D E
F G
1 0.0050 0.0432 0.0877 0.6400 0.1310 0.0931 0.0
~
2 0.0035 0.0576 0.0707 0.5891 0.1341 0.1449 0.0 3
0.0099 0.1179 0.0977 0.4585 0.1054 0.2105 0.0 4
0.0044 0.1259 0.0866 0.5346 0 0804 0.1679 0.0 5
0.0145 0.1249 0.1293 0.4855 0.0702 0.1756 0.0 6
0.0082 0.1423 0.1735 0.4281 0.0437 0.2043 0.0 7
0.0114 0.0972 0.1352 0.5262 0.0332 0.1968 0.0 8
0.0013 0.0731 0.1359 0.5747 0.0441 0.1710 0.0 9
0.0027 0.0465 0.0840 0.6252 0.0848 0.1587 0.0 10 0.0023 0.0442 0.0773 0.5198 0.1294 0.2269 0.0 11 0.0014 0.0321 0.0640 0.5891 0.1424 0.1709 0.0 12 0.0025 0.0241 0.078S 0.5368 0.1460 0.2121 0.0 13 0.0068 0.0426 0.0922 0.4321 0.1528 0.2734 0.0 14 0.0060 0.0476 0.0945 0.4641 0.1209 0.2669 0.0 15 0.0078 0.0537 0.1385 0.5196 0.1616 0.1188 0.0 16 0.0078 0.0554 0.1399 0.5629 0.1192 0.1148 0.0 J
l l
i 6
i A-11
~~
Table A.8. Frequencies of wind directions and true-everage wind speeds Wind speeds for each stability class Wind toward Frequency (m/s)
A B
C D
E F
G N
0.154 2.20 2.50 4.10 4.90 3.50 2.10 0.0 NNW 0.061 2.10 2.40 3.80 4.20 2.90 2.10 0.0 NW 0.029 1.80 2.10 2.90 2.90
- 2.20 2.60 0.0 WNW 0.016 1.00 2.10 2.40 3.30 2.80 1.50 0.0 W
0.015 2.20 2.00 2.50 3.50 2.70 1.40 0.0 WSW 0.017 2.30 2.20 3.00 3.50 2.80 1.60 0.0 SW 0.031 1.80 2.40 3.00 3.80 2.50 1.60 0.0 SSW 0.053 1.50 2.20 3.10 4.00 2.90 1.60 0.0 S
0.156 2.20 2.20 3.20 4.50 3.10 1.90 0.0
.l SSE 0.093 2.20 2.40 3.80 5.00 3.40 2.00 0.0 SE 0.100 2.10 2.20 3.80 6.30 3.70 2.00 0.0 ESE 0.084 2.10 2.00 4.30 6.30 3.80 1.90 0.0 E
0.062 2.20 2.40 3.70 5.90 3.30 1.90 0.0 ENE 0.047 2.50 2.30 3.60 5.30 3.10 2.10 0.0 NE 0.036 2.20 2.60 3.90 6.30 3.40 2.30 0.0 NNE 0.045 2.10 2.50 4.20 5.10 3.40 2.10 0.0 i
Table A.7. Other parameters used in determining exposure to air I
concentrations of radionuclides released in the building vent effluents Parameters Quantity or dimensions Stack height 9m Stack, diameter 0.51 m Stack sverage air How 0.94 m/s Temperature fannual average for Jreal 10"C Rainfad (annua 6 averages 108 cm/ year Heqht of led (annual average) 765 m i
i
f e
A-12 Table A.S. Popuestion dem for 1980 at Windsor Locks -latrtude = 41.882217, lonytude = 72.18887 Distance, miles incrementes population date 0.0-1.0 1.0-2.0 2.0-10 3.0-4.0 4.0-5.0 5.0-10.0 10.0-20.0 20.0-30.0 30.0 40.0 40.0 -50.0 N
8 8
8 586 8
2578 48873 45248 26499 27398 NME 8
196 325 8
38 4897 188719 138518 42163 9146 NE 8
8 344 1919 1889 16182 54488 39142 28336 28887 ENE 8
198 448 828 4864 18765 14832 18122 26494 56152 E
8 8
747 758 1325 4829 36933 17593
?647 35788 E5E a
398 8
8 990 6785 36898 48171 18535 31282 SE 289 8
569 1516 156J 28742 48872 12135 17987 121898 5
g SSE v
541 8
2586 3
88735 58885 27181 28773 36858
=Q 5
6 8
9 4743 6282 04185 109774 87331 47918 5e775 SSW 6
461 8
738 a
21298 96733 182869 144613 678569
$U 249 8
472 Sl9 628 4887 62597 112355 34780 69638 WSu 6
274 a
a 452 7tsa 15348 36627 18796 25468 W
8 8
259 455 2855 3273 14197 7371 7686 14917 iJiu a
535 452 877 435 1757 18833 3578 8919 7161 Nu d
J 731 539 389 2355 2047 1274 8923 47123 wJ d
522 563 8
3tf 2144 8762 3692 3049 38251
'DTAL 458 3169 4982 15879 29147 273684 791823 676391 446971 126?d23 Cumulative populatron data 0.0-1.0 0.0-2.0 0.0-3.0 0.0-4.0 0.0-5.0 0.0-10.0 0.0- 20.0 0.0-30.0 0.0-40.0 0.0 - 50.0 1
6 8
8 586
$86 3876 43949 89189
!!!688 143886
- E a
196 521 521
- 59 4656 185375 315885 358848 367194 NE 8
4 344 2263 4152 28334 74742 113894 134228 154227 ENE 4
198 638 1458 5522 16237 3 tit 0 41241 67735 123887 E
4 3
747 1585
?833 7659 14592 62165 71832 187628 EIE 6
398
!?3 398 128'J 7985 24883 35854 1035??
134971 il 289 289 778 2294 3657 24599 72671 91806 102713 224611 E3 ISE 8
541 541 7847 3847 93792 142587 169758 198546 227484 d
5 4
8 1
4748 13953 95135 2049R9 29224n 340159 3*G933 ISU J
461 41 1t99 1199 22439 119222 2212?!
165904 181.u?3 SU 249 249
'21 1249 1068 6755 69352 181737 216415 29fe45 JSU 274 274 274 726 7994 23?34 53961 70657 97!25 4
3 8
259 714 2764 6037 20234 2?G05 352*1 5A208 WJ d
591 1847 1924 2359 4116 14149 17727 25E43 32007 N
a G
721 12da 1549 3984 5351 7225 16140 53271
-NJ
$22 1895 IJ85
- 402 33 4 123N 1p00 19041 421*?
T QL 459 3627 3!28 24M9 44!?1 34254 110927'
'79*ff4 22:4639 250 0 52 y --
,<c,
a A-13 REFERENCES FOR APPENDIX A 1.
R. E. Moore et al., AIRDOS-EPA: A Computerized Methodology for Estimating Environmental Concentrations and Dose to Man from Airborne Releasse of Radionuclides, ORNL-5532, Oak Ridge National Laboratory, Oak Ridge, Tenn., June 1979.
2.
F. Pasquill, 'The Estimation of the Dispersion of Windborne Material,"
Meteorol. Mag. 90:33 (1961).
3.
F. A. Gifford, Jr., "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," Juct. Saf. 2(4):45-57(1961).
4.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, Calculation of Annual Doses to Man frcm Routine Releases of Reac or Effluents for the Purpose of Evaluating Compliance m*th 10 CFR Part 50, Appendix I, Rev.1, Office of Standards Development, Washington, D.C. (1977).
5.
J. C. Pleasant, INGDOS - A Conventional Computer Code to hpiement U.S. Nuc2 ear Regulatory Guide 1.109 Modeis for Estimating the Annuai Doses from Ingestion of Atmospherically Released Radionuclides in Food, ORNL/TM-6100, Oak Ridge National Laboratory, Oak Ridge, Tenn., 1979.
6.
D. C. Kocher, Dose-Rate Conversion Factors for F ter~.at Eaposure to Photons and Electrons, ORNL/NUREG-79, Oak Ridge Nation &1 Labora';ory, Oak Ridge, Tenn., August 1981.
7.
D. E. Dunning, Jr., S. R. Bernard, P. J. Walsh, G. G. Killough, and J. C. Pleasant, Estimates of Inter".at Dose Ecuivalent to 22 Target Crgans for Radionuclides Occurring in Rou:ine Releases frcm helear hel-Cycle Facilities, vol. II, ORNL/NUREG/TM-190/V2, Oak Ridge National Laboratory, Oak Ridge, Tenn., October 1979.
8.
ICRP Task Group on Lung Dynamics, "Decosition and Retention Models for Internal Dosimetry of the Human Respiratory Tract," Eea!:h Phys.12:173-207 (1966).
9.
I. G. Eue, "A Review of the Physiology of the Gastrointestinal Tract in Relation to Radiation Doses from Radioactive Materials," Esa!!h Phys.12:131-162 (1966).
10.
G. G. Killough and L. R. McKay, eds., A Me:hodology for Calculating Radia: ion Doses frcn Radioac:ivi:y Released : :he Envircr:en:,
ORNL 4992, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 1976.
A-14 11.
D. L. Shaeffer and E. L. Etnier, AQUANA3 - A computer Code for Calculating Dose Ccmmitments to Man from Aqueous Releases of Radionuclides, ORNL/TM-6618, Oak Ridge National Laboratory, Oak Ridge, Tenn., February 1979.
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