ML20202G135

Lynchberg Research Ctr Environ Rept

ML20202G135
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Site:	07000824
Issue date:	06/30/1986
From:	BABCOCK & WILCOX CO.
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June, 1986 I

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I LYNCHBURG RESEARCH CENTER ENVIRONMENTAL REPORT I

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Babcc + % Wilcox Co.

I Lynchburg Research Center Post Office Box 11165 Lynchburg, Virginia 24506-1165 I

l 8607150309 860623 l

PDR ADOCK 07000824 C

PDR

I June, 1986 I

TABLE OF CONTENTS r

Section Page 1.0 PROPOSED ACTION.

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2.0 THE SITE 2-1 3.0 THE FACILITY.

3-1 4.0 ENVIRONMENTAL EFFECTS OF SITE PREPARATION AND PLANT CONSTRUCTION AND OPERATION 4-1 5.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS 5-1 6.0 EFFLUENT AND ENVIRONMENTAL MEASUREMENTS.

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I June, 1986 1.0 PROPOSED ACTION This environmental report is submitted by the Babcock & Wilcox I

Company, Lynchburg Research Center to the U. S. Nuclear Regulatory Commission in support of its request for renewal of License SNM-778.

This action is taken pursuant to Title 10, Code of Federal Regu-lations, Part 51.

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1.1 BACKGROUND

INFORMATION Babcock & Wilcox, an operating unit of McDermott, Inc., is a major industrial company which manufactures and markets specially engineered industrial products and naterials which help perform essential tasks I

for utilities, industries, institutions and governments throughout the worl d.

I More than half of the unit's business is in the design, manufacture and erection of energy systems and components. The balance is in specialty steel tubing, refractories, advanced composites, automated machinery, valves and process controls.

The production of steam supply systems has traditionally been a major part of B&W's business. Today, B&W is one of the leaders in the I

design and production of both nuclear and fossil power generating equipment for utilities, industry, ships, schools and hospitals.

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The Research and Development Division provides the operating divi-sions of B&W with the technical leadership and skills necessary to develop new products and processes, and to examine and improve those of the present generation. The R&D Division is comprised of two I

research centers. One is located in Alliance, Ohio, which is also the division headquarters and the other in Lynchburg, Virginia which tnis report addresses.

The Lynchburg Research Center (LRC) was first known as the Critical Experiment Laboratory when it began operation in 1956 as a part of the Atonic Energy Division.

In 1957 the AEC issued License CX-1 for the I

operation of the first privately owned and operated critical experi-ment facility in the United States, which was located at the Laboratory. This facility was used to design and test the first I

nuclear core for the Consolidated Edison Power reactor. This was a thoriun core, the first of its kind built in the U. S.

In 1958, additions to the Critical Experiment Laboratory included I

facilities for the nuclear merchant ship critical experiment, the Lynchburg Source Reactor (CX-12), and the Lynchburg Pool Reactor (R-47).

I The Laboratory expanded again in 1964 with the addition of the Nuclear Fuels Laboratory. This building included the Babcock and Wilcox Test I

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June, 1986 Reactor (BAWTR), an oxide fuel preparation laboratory, and a hot cell i

facility. At this time the laboratory name was changed to the Nuclear Development Center.

In 1966, the Nuclear Development Center became a part of the Research and Development Division and its present name was adopted.

In 1968, the Plutonium Development Laboratory was added. This facility was l

I built to accommodate the equipment necessary to accommodate plutonium l

s mixed oxide fuel preparation and examination.

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I The LRC presently employes 180 scientists, engineers, technicians and support personnel. Approximately 30% of the work is performed under NRC licenses. The remainder is in the areas of process control, non-destructive examination methods and instrument development, and non-I nuclear ceramics.

Research and development utilizing source, byproduct and special I

nuclear material is performed primarily in one building. Building A is presently being decommissioned.

Ruilding B houses the hot cell facility, the crane and cask handling I

area, a radiochemistry laboratory, a counting laboratory, and a scanning electron nicroscope laboratory. The four hot cells are used to handle and examine materials that are highly radioactive.

Irradi-I ated commercial nuclear fuel assemblies have been partially dis-assembled and destructive and nondestructive examinations perforned on the fuel rods. Reactor irradiated experiment capsules have been I

disassembled and studied and examinations of primary system components are performed. The cask handling area, the radiochemistry laboratory and the scanning electron microscope laboratory support the hot cell operations.

I Ruilding C is presently being deconnissioned.

I The Radioactive Waste Storage Building is used to house containerized radicactive solid waste subsequent to shipping for off-site disposal.

The Liquid Waste Disposal Facility is a " tank farn" where process area liquid wastes are collected, stored, sampled, diluted, and pumped to the waste disposal facility of the Naval Nuclear Fuel Division.

I 1.2 REGIONAL SITE LOCATION The selection of the Mt. Athos site n the location of the najor I

portion of B&W's nuclear fuel activities was based on a number of criteria related to social, environnental and economic factors. The validity of the choice has been demonstrated by the successful conduct of nuclear-related activities at the site since 1956 t

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I June, 1986 1.2.1 Social Factors The Lynchburg area possesses an exceptionally stable and productive jI work force. Additionally, the acceptance of nuclear activit.ies as a safe and valuable industry has resulted in excellent community re-lations with B&W. The company is not aware of any adverse reaction to the location of the facility on the outskirts of Lynchburbginia, Lynchburg, which is one of the major industrial centers in V t

has a solid and varied industrial base, a pleasant climate, active community organizations, and plentiful recreational opportunities.

I 1.2.2 Economic Factors The LRC provides its services principally for the company's operating division. Three of its customers; the Nuclear Materials Division, the Naval Nuclear Fuel Division, and the Nuclear Power I

Division are located in the Lynchburg area. Other company facilities extensively utilizing LRC services are located in Georgia, Ohio and Pennsylvania. The LRC is therefore centrally located in relation to its principal customers.

The Lynchburg area is serviced by two major railroad systems, several connercial airlines, and a number of major trucking firms.

Campbell County and the Commonwealth of Virginia possess favorable tax structures for industry. Also, the stability of the local work force contributes significantly to productivity while reducing the I

economic penalty associated with a large labor turnover.

1.2.3 Environmental Considerations The LRC is located approximately four miles from the nearest major I

population concentration and occupies approximately 525 acres of land formerly devoted to agricultural pursuits. The site itself was selected after geological and hydrological investigations had de-termined its acceptability for nuclear activities on the basis of I

geological stability, groundwater flow characteristics, and hydro-logical considerations relating to the neighboring James River. The maximun flood crest recorded for the James was approximately 100 I

feet below the LRC. Additionally, the Center is only minimally affected by storms noving inland from the Atlantic Ocean and the hilly nature of the surrounding country causes meteorological conditions to be relatively stable.

1.3 PROPOSED PROJECT SCHEDULE The proposed project under consideration is license renewal and there-fore this section is not applicable.

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June, 1986 1.4 PREVIOUS ACTION ON APPLICATION Table 1.1 provides a detailed history of AEC and NRC licensing t

activities relating the the LRC.

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June, 1986 3.0 THE FACILITY 3.1 EXTERNAL APPEARANCE The buildings that comprise the LRC are all of masonry construction.

3.1.1 Building A t

Building A is constructed of concrete block basically. The walls of the critical experiment bays are pcured concrete. That portion of the building which faces the Naval Nuclear Fuel Division (SSE),

has a red brick facade. All of the windows except those in the east corner and south second floor are solid pane, vertical rectangles. The exceptions are multipane, horizontal rectangles.

3.1.2 Building B Building B is a two-story structure.

It is constructed of concrete block with a gray agrigate brick facade on the south face. A series of seven vertical rectangular projections are located on the south central face. Six of the projections con-tain first and second floor vertical rectangular windows accented at the top and bottom by green stone slabs. The seventh contains a second floor window and the front door. The building is 340 I

feet by 98 feet. The south lawn is landscaped with evergreen shrubs and hemlock trees. The remainder of the building is surrounded by a lawn of grass.

3.1.3 Building C Building C is a single-story building of concrete block con-struction. Outside dimensions are 225 feet by 174 feet at its deepest point. The front of the building faces south. The right-hand side of that face contains the the eight windows of the building and its front door. The block face is covered with painted stucco. A driveway abu*tts the front left-hand portion of the front of the building and the read right-hand portion. The front right-hand portion is a grassy lawn with evergreen scrub landscaping.

Those areas of t' e building not abutted by the r

driveway as grassy lawns.

3.1.4 Building J Building J is the solid waste storage facility.

It is located in the rear of Building C.

This building is a single-floor concrete I

block square structure. The building contains no windows. A e

single personnel entrance and a large roll-away door are located on the south face and a large roll-away door is located on the I

north face. The building exterior is painted a beige. The building is surrounded by asphalt paving and a chain link fence.

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3 June, 1986 3.1.5 Liquid Waste Disposal Facility The Liquid Waste Disposal Facility is located to the southeast of Building C.

It is a single-story concrete block building covered with stucco and painted beige. This building has a single personnel entrance door on the north face and a double door on the south. Grassy lawn abutts the building on the north, west and south sides and a concrete slab abutts it on the east.

s 3.1.6 Building D Building D is a complex of six buildings. Five of these are single-floor, concrc te block buildings with grey agrigate brick facing on all sides. The central building is two stories high I

with a grey agrigate brick facing on three sides and red brick facing on the front or west face on the first floor. The facing of rock agrigate panels on the second floor accents the 23 single-I pane vertical rectangular windows and overhangs the first floor entrance. This complex is landscaped with evergreen scrubs, small hardwood trees and evergreen trees on the west lawn. The remaining sides are abutted with grassy lawn.

3.2 PLANT OPERATION Operations at the LRC are widely diverse and change frequently. A brief description of current operations is given in the sections that foll ow.

Due to the frequent changes in work performed in specific labora-tories, resources used and effluents for the LRC are presently in the following tables:

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June, 1986 TABLE 3.1 AVERAGE MONTHLY UTILITY USAGE, 1984 I

Water 580.000 gal.

Natural Gas 600,000 cu.ft.

i Electricity 400,000 kwh I

TABLE 3.2 TOTAL ANNUAL NON-RADI0 ACTIVE EFFLUENTS 1984 Ai r 2 x 1010 ft3 Sanitary Sewage 7 x 106 gal.

Solid Waste (Trash) 6 x 104 ft3 I

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I June, 1986 TABLE 3.3 RADI0 ACTIVE LIOUID WASTE RELEASES TO NNFD TREATMENT SYSTEM I

1982 TO 1984 I

1 Microcuries Nuclide 1982 1983 1984 Mn-54 1.8 Co-58 9.6 11.0 Co-60 100.0 30.0 390.0 I

Sr-90 49.0 4.6 44.0 Y-90 49.0 4.6 44.0 Sb-124 15.0 I

Sb-125 35.0 Cs-134 14.0 37.0 0

Cs-137 140.0 980.0 840.0 Ce-144 41.0 Gross Beta 170.0 110.0 350.0 Gross Alpha 56.0 42.0 100.0 Pu-241 220.0 Plutonium 280.0 Totals 1100 1200 1900 (rounded)

Total Volune, 294.000 180,000 384,000 gal.

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June, 1986 TABLE 3.4 GROSS RADI0 ACTIVITY RELEASED FROM 50 METER STACK I

1982 TO 1984 1982 1983 1984 Gross Alpha Particulate 0.2 pCi 0.2 pCi 0.3 pCi 1

Gross Beta Particulate 2.3 pCi 1.5 pCi 2.7 pCi s

Kr-85 0.09 Ci 0.09 C1 14.0 Ci Tritiu,(3) 0.007 C1 0.0007 Ci 1.1 Ci NOTES:

1.

Stack flow is 25,000 cfm 2.

LLD's for stack nonitor or stack sampling Kr-85 6 x 10-7 pCi/ml Gross Alpha 2 x 10-16a pCi/ml Gross Beta 6 x 10-168 pCi/ml 3.

Tritium release is calculated based on opening I

tritium containing reactor components in the hot cell.

4 The stack is sample continuously at a nominal I

sample rate of approximately 2.5 cfm (3.7E10 ml/yr) with a minimun sample rate of 2.0 cfm. Samples are 1

analyzed weekly for gross Beta and Alpha.

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June, 1986 TABLE 3.5 SOLID RADI0 ACTIVE WASTE 1984 I

Nuclide Quantity Byproduct Co-60 0.41 millicuries Sr-90 1.8 millicuries Y-90 1.8 millicuries I

Cs-134 1.2 nillicuries Cs-137 15.0 millicuries An-241 0.81 millicuries Fissile and Pu U-235 0.0377 mci Pu-238 0.017 mci Pu-239 3.07 mci Pu-240 0.088 mci I

Pu-241 0.515 mci Pu-242 0.16 pCi Source Material I

t U-234 0.432 mci U-238 0.150 mci Th-232 0.021 mci Total volume of waste:

2200 cu.ft.

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June, 1986 3.2.1 Building A (Figures 3-1 and 3-2)

Building A is in the process of being decommissioned. All account-able quantities of licensed naterial have been removed from the building. The only work in process under the licenses (CX-10 &

SNM-778) is decommissioning.

3.2.2 Building B (Figures 3-3 and 3-4)

This facility is comprised of four hot cells, a hot cell operations area, a cask handling area, a transfer canal and storage pool, a hot machine shop for work on contaminated equipment, an experimental pool, radiochemistry laboratory, two netallurgy laboratories, a counting laboratory, a health physics counting area, a ceramics oven I

roon, a machine shop, a failure analysis laboratory, a scanning electron microscopy laboratory and a fatigue & fracture laboratory.

Radioactive solid, liquid and gaseous releases are combined in the totals for the Center, as is the nonradioactive effluents. Figures 3-5 and 3-6 show the facility ventilation and liquid waste systens.

l 3.2.2.1 Hot Cell Facilities This facility consists of four hot cells, an operations area, the cask handling area, the transfer canal and storage pool and the hot machine shop.

The transfer canal and storage pool is used to receive, unload, I

load and prepare shielded casks for shipment.

It also is used for storage of radioactive material and for transferring radioactive material to and from the hot cells. The pool water is recircu-lated through ion exchange columns for cleanup. These resins are I

replaced when expended and handled as dry waste. Particulates that settle to the pool bottom are removed periodically with an underwater vacuum cleaner and disposed of as dry waste.

The hot cells are used to perforn destructive and nondestructive testing and exanination of highly radioactive materials. These include reactor core hardware components and fuel rods renoved j

from irradiated reactor fuel assemblies. The cells generate solid and gaseous radioactive wastes. The gaseous wastes consist of Krypton which originates from irradiated fuel rods that are I

punctured for fission gas analysis. The iodine component has decayed prior to shipment to the LRC. The estinated Krypton r'e-lease over the 5 year period beginning with 1986 is 50 Curies.

Solid waste in the form of irradiated fuel is placed in double containers and placed in interin storage in the Waste Storage Tubes that are located in the Cask Handling Area or adjacent to the Liquid Waste Building. These are 6-inch diameter steel tubes that are enersed in concrete. Leakage from this configuration is not credible and the tubes are not tested for leakage. This form I

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I June,1986 of waste is addressed in the Nuclear Waste Policy Act of 1982, and will be disposed of by the Department of Energy.

Solid waste that is generated in the hot cells that contains I

transuranium elements is placed in steel drums and placed in interin storage in the Annex to Building J.

This waste is addressed in the Nuclear Waste Policy Act of 1982, or is addressed in contracts with that department and in both cases will be dis-s posed of by the Department of Energy. The estimated volume of this fom of waste for the 5 year period beginning in 1986 is 305 I

cubic feet. Other solid waste containing byproduct naterials are stored in appropriate shipping containers in Building J or in the area surrounding it.

This waste is periodically sent to a com-nercial waste disposal facility pursuant to 10 CFR 61.

The hot nachine shop is used when repair of manipulators is required and for perfoming work on items that are radioactive but not to the extent that remote hot cell handling is required.

Solid radioactive waste is generated in the area.

The cask handling area is a high bay room used to receive and ship I

containers of radioactive material. The largest source of waste is generated in decontaninating shipping containers. Liquid waste in the fom of scrub water is released to the liquid waste retention basins.

The operations area contains the manipulator operating stations, the fission gas analyzer and the electronic equipment associated I

with the nondestructive analyzers. No radioactive wastes are generated in this area. Nonradioactive solid waste is included in the total for the LRC.

3.2.2.2 Experimental Pool This 30,000 gallon pool is used to develop underwater exanination I

equipment. Radioactive material is not now handled in this pool.

3.2.2.3 Radioch2mistry Laboratory This laboratory utilizes standard chemical fume ho'ods the exhausts of which pass through one prefilter and one HEPA filter.

Work of interest being presently perfomed is analysis of irradiated fuel sanples, corrosion products, neutron flux dosi-meters and reactor coolant samples. Low level radioactive wastes are released through the liquid waste disposal facility. Other liquid wastes are solidified for off site burial. Solid waste is shipped for off site burial. Airborne and gaseous effluents are filtered and discharged through the 50 mete. exhaust stack. All I

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June, 1986 these contributions are included in the site totals.

3.2.2.4 Metallurgy Laboratory The metallurgy laboratory has equipment for structural exani-nations on a macroscopic and microscopic scale. Facilities are available for all metallography preparations and examinations I

utilizing light-microscopy. A hot stage metallograph is avail-1 able for microscopic examination of materials at high temperatures and in controlled atmospheres. An industrial x-ray unit is also available to this laboratory.

Wastes from the metallurgy laboratory are cypically nonradioactive and solid. Water used for cooling is discharged to the storm I

drains, 3.2.2.5 Counting Laboratory The counting laboratory contains several high resolution gama spectroscopy systens coupled to computers for data processing. A liquid scintillation systen is used for spectrosccpy of low energy beta enitters. Gross counting and spectroscopy are performed on alpha and beta enitting elements.

_.I The laboratory is not equipped with sample preparation facilities.

Preparation is performed in other laboratories and transferred to the counting laboratory and returned after counting to the sending l aboratory. No releases are made from this laboratory.

3.2.2.6 Ceranics Oven Roon I

This roon is used for nixing, forming and sintering nonradioactive ceramic nat cials.

I Wastes are primarily solids that are included in the LRC solid waste totals. Cooling water is discharged to the storn sewer.

3.2.2.7 Scanning Electron Microscopy Laboratory Radioactive and nonradioactive specimens are prepared and exanined in this facility. Small amounts of solid wastes are generated and these are included in the L'lC totals.

3.2.2.8 Fatigue A Fracture Laboratory This laboratory contains a closed-loop electrohydraulic load frane, and impact tester and a fatigue precracker.

Specimens are brought into this laboratory for testing and lI l

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l returned to the originating laboratory for disposal, l

3.2.2.9 Failure Analysis Laboratory l

This laboratory is equipped for perforning exanination and testing of components that have been removed from nuclear power plants.

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Examination and testing includes visual, photography, dimensional 1

measurement, metallographic preparation and exanination, and l

corrosion testing. Small amounts of solid and liquid wastes are generated and are included in the site totals. Ventilation is I

provided by filtered fune hood off-gas which draws air from the cask handling area.

3.2.3 Building C (Figure 3-7)

Building C is in the process of decommissioning. All accountable quantities of licensed material have been removed from the building.

The only licensed activity in the building is decommissioning.

3.3 WASTE CONFINEMENT AND EFFLUENT CONTROL 3.3.1 Air Effluents The exhaust air from the LRC is made up of two streans, air exhausted from hoods, glove boxes, bot cells, and potentially con-taminated areas, and general building air necessary to maintain confort.

Exhaust air from hoods, glove boxes and hot cells is passed through a prefilter and at least one stage of HEPA filtration prior to release via the 50 meter high stock (Figure 3-5). Room off-gas from areas where there exists the potential for airborne radioactive contanination is passed through a prefilter and one stage of HEPA

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filters and is released through vents at essentially roof height.

General building air is partially recirculated for energy conserva-l tion and released at roof height.

3.3.1.1 Controlled Area Air Effluents Exhausts from hot cells, fune hoods and glove boxes are the main j

sources of supply to the 50 neter high stack. This stack is sampled isokinetically continually while work in these areas is in i

progress. A drawing of the systen is shown in Figure 3-5.

Air I

passing into the stack has been filtered through at least one

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I stage of HEPA filters.

In the case of the hot cells, glove boxes and Building C fune hoods, two series stages are used. One perchloric acid fume hood presently installed, is an exception to the above practice. This hood exhausts directly to the roof of Building B with no filtration.

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June, 1986 Releases through the 50 meter stack are given in Table 3.4 3.3.1.2 tbnradioactive Effluents The nature of the work performed at the LRC is such that only small amounts of volatile chemicals are used. The single largest contributor is acetone, of which the Center consumed 100 gallons in 1983. On the basis that 100 percent of the material evapo-rated and was released through the ventilation systen 1.82 lb/ day would result.

3.3.2 Liquid Effluents Liquid effluents leave the LRC by three routes; the storm sewer which not onl; carries rain water but the major portion of cooling water, the sanitary sewage line which flows to the treatment facility at tne Naval Nuclear Fuel Division (NNFD) and the only noteworthy one of the three, the effluent from the liquid waste retention tanks which flows 11to the treatment facility at NNFD.

3.3.2.1 Contaminated Liquid Waste System Potentially contaninated and contaminated liquid wastes fron laboratories are directed to the liquid waste dispnsal system.

The drain systen for Building B is shown in Figure 3-6.

A l

schenatic diagran of the liquid waste retention tansk and piping is shown in Figure 3-8.

The liquid waste retention basins are l

monitored for leakage by observing decreases in the liquid levels.

There has been no evidence of leakage from any of the basins since their installation.

All waite tanks are sampled quarterly and prior to emptying.

Prioreto sampling, a tank is thoroughly mixed. A sanple is taken by di.ging a clean container into the tank through a manhole at the top. From the sanple, 5 ml is evaporated to dryness and counted to determine the gross Alpha and Beta activities. The re-maining sample is Ganma Scanned for isotopic identification. com-parison of the gross Beta and Gamma scan results may indicate the presence of Sr-90, in which case a second sample is drawn and sent to a contractor laboratory for Sr-90 analysis. The analyses must confinn that the concentration of radioactive naterial in the tank I

liquid meets the 10 CFR 20, Appendix B, Table II, values for release to an unrestricted area, prior to release to NNFD.

If f

thses limits are exceeded, dilution is used to bring the concen-i tration into compliance. A compilation of releases through this i

system is presented in Tab 1? 3.3.

3-11

~

June, 1986 3.3.2.2 Sanitary Waste Effluents Untreated sanitary wastes are combined for treatmertt with those of the NNFD's at that facility's sanitary waste treatment facility.

4 The LRC's contribution to this facility is 1.9 x 10 gallons per day. No radioactive material is discharged through the sanitary sewage system.

3.3.2.3 Storm Drainage Runoff from the parking lot, building roofs and surrounding land is collected by the storn drain system. Water used for cooling furnaces and similar uses is also collected by this system. The system discharges into a pond that is located on the east side of the LRC, to the rear of Building J (Figure 2-6).

The overflow from the pond discharges to a dry strean bed and then it flows to l

the James River. The pond is sampled monthly. Radioactive material is not discharged to the storm drain systen, i

3.3.3 Solid Wastes All solid wastes generated from LRC operations are nonitored and disposed of as described below.

3.3.3.1 Contan'nated Solid Wastes Contaminated solid wastes are disposed of by a NRC-licensed facility. These wastes consist of filters, packing material, I

decontanination equipment, contaminated laboratory equipment and solidified liquids. These wastes are packaged and stored at the LRC until a sufficient amount has accumulated for shipnent to bu ri al. Packaged wastes are stored in a building specified for this purpose. A fenced area adjacent to this building is used for storage of packaged LSA and fissile exempt material.

3.3.3.2 Uncontaminated Solid Wastes s

4 Approximately 6 x 10 cubic feet of uncontaninated solid wastes are generated at the LRC per year. These wastes are routinely monitored to ensure that they are not radiologically contaninated and disposed of by a private contractor at the Lynchburg sanitary landfill. Salvageable materials, such as netals, are sold or recycled.

3-12

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June, 1986 I

4.0 ENVIRON'iENTAL EFFECTS OF SITE PREPARATION AND PLANT CONSTRUCTION AND UPERJ ION _

4.1 EFFECTS OF SITE PREPARATION AND CONSTRUCTION The facility, for which a renewal of the operating license is being sought, is already in existence. At this time, the major structures are completed and in operation, and the unused land has been graded and landscaped or allowed to return to its natural state.

Any social, economic, or ecological impacts of construction are now history. Undoubtedly, construction aided the economy of the region by providing employment, and presumably small areas of land were renoved from biological productivity and dedicated to research and development I

use. There is no evidence to indicate that the social, economic, or ecological impacts of construction were harmful, or even of a very great magnitude.

4.1.1 Land Use The site was originally used for farming and consequently much of I

the land had already been cleared long before construction of the LRC. Since cessation of farming, much of the land has been retaken by shrubs and trees.

In addition, areas that were denuded of vege-tation before construction have been reforested with pine trees and grassy neadows.

The natural landscape has been altered to acconnodate buildings, I

parking lots and access.sads.

Each of these has been designated to ninimize undesirable environmental effects. Overall, these altera-tions have not had an adverse effect on terrestrial life. Si gni fi-I cant portions of the site remain suitable for plants and wildlife species. No observable erosion, dust, or excessive noise caused by traffic or plant operation is evident.

4.1.2 Water Use Changes in the contour of the land that were required for con-struction of parking lots, roadways, and buildings did not signifi-cantly alter the natural drainage patterns of surface water flow.

4.2 EFFECTS OF PLANT OPERATIONS 4.2.1 Radiological Impact 4-1 I

June, 1986 4.2.1.1 Airborne Effluents As stated in the note in Table 3.4, air releases from operations at the LRC are in all probability attributable entirely to back-ground for long-lived particulate activity. During irradiated I

fuel examinations, Kr-85 is released. Table 3.4 shows that about 14 Ci of Kr-85 may be released as a result of these operations in a year. The Kr release in 1984 was higher than the two previous years shown in Table 3.4 because irradiated fuel was examined in s

the hot cell in that year. The estimated release per year for the next five years, beginning in 1986, is 10 Ci. The exposure to a person living in the City of Lynchburg from this type of release is 0.00007 man-rems assuming the following:

o The variability of the winds to Lynchburg which is due west of the LRC is 4.5%.

o The City of Lynchburg is five miles west of the LRC.

o The population of Lynchburg is 60,000 people, o An individual exposed to an integrated cloud of 1 Ci sec/m3 = an exposure of 7 x 10-4 Ren.

4.2.1.2 Liquid Effluents Referring to Table 3.3, during the period July through necember, 1983, 0.98 millicuries of Cs-137 was released to the NNFD waste treatment systen. This is the highest release of the period I

covered.

The nan-rem exposure for a release of 0.98 nillicuries of Cs-137 per year to the James River is given below.

1.

The nearest munipical water systen utilizing the Janes River, downstrean of the LRC, is the city of Richmond.

2.

The 1970 census gives the population of Richmond as 250,000.

3.

The average flow rate of the James River is 5500 ft3 per

second, o

Concentration at Richmond:

0.98 x 103 pCi (5500 ft3/sec)(365 days)(24 hr/da)(3600 sec/hr)(2.83 x 104 ml/ft3) 2 x 10-13 pCi/nl.

=

4-2

5 I

June, 1986 I

o An adult drinks 370 liters of water per year, o

The dose ingestion conversion factor is 7.14 x 10-5 mRen/pCi ingested.

o Activity ingested:

I (1.02 x 10-12 pCi/ml)(3.70 x 105 ml) = 7.39 x 10-8 pCi o

Man-ren exposure in Richmond is:

(0.074 pCi)(7.14 x 10-5 mRen/pCi)(250,000 people)

I 1.3 nan-milliren.

=

4.2. 9.

Chemical Discharge I

Chemical discharges from the LRC are made through the liquid waste disposal systen. Based on the receipts of hazardous chemicals, less than a liter per day is discharged. These discharges are made to the NNFD liquid waste treatment system and is included in that I

f acility's sample results.

4.3 RESOURCES COMMITTED The following res'ources were conmitted for the facility:

1.

The land 2.

Structural naterials 3.

Nonrecoverable consunables used during construction.

The only land permanently affected by construction was the four acres I

enclosed by the security fence of which 2.25 acres is occupied by buildings, and 2.1 acres in parking lots and driveways. Since the site is in an unpopulated rural area with abundant unoccupied land, I

the dedication of 6.1 acres for the facility has no noticeable effect on the natural ecosysten, and it does not permanently foreclose other options for human development.

I I

4-3

I June, 1986 I

4.4 DECOMMISSIONING AND DISMANTLING

Reference:

Babcock & Wilcox I

Research and Development Division Lynchburg Research Center Renewal Application i

I License SNet-778 October, 1985 I

I I

I I

I I

1 I

!I

I

,I 4-4

I June, 1986 I

5.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS 5.1 GENERAL Several accidents have been postulated and analyzed for the LRC. Some of these are unique to our type of operation and do not fall into the categories normally considered for a fuel fabrication facility.

I 1

5.1.1 Power Failure, Hot Cell I

A potential hazard would be total utility power failure to the LRC site, along with failure of the standby engine to start.

It is standard practice to secure all hot cell operations in a safe manner whenever an LRC power failure occurs.

In this assumed situation, I

the Hot Cell ventilation is maintained by one fan connected to the emergency bus. One fan is adequate to maintain a delta P of 0.25 inch of water over the cell face. However, emergency power from the I

notor generator is provided to both fans, and this emergency power source is checked once a week to ensure startup. This notor generator is equipped with an automatic starting nechanism and a backup manual starter if the automatic starter should fall. Hot I

cell emergency-lighting is sufficient to permit limited operations to safely secure the cell.

I Ventilation is maintained through the normal duct work, which con-tains a prefilter and absolute filters that remove particulate cateri al s. The Hot Cell operations that produce radioactive gases are handled in such a way that these gases are contained. Fission I

gases can only be released to the cell atmosphere by manual oper-ation of a valve. The gas is released only after an estinate of gross activity is complete. Gas release nf this type is allowed I

only during nornal ventilation conditions and is stopped immedi-ately in the event of an energency. Tnus, it is apparent that, even with a complete loss of power to the facility, the surrounding area is adequately protected.

H :.;,t V 1 ventilation air joins that from the clean areas in the nanhole oehind the main building. Failure of any number of fans in other parts of the system would not cause a backup into that portion of the systen, since the suction from the stack fan would provide an air velocity greater than 100 fpn from the manhole. Backdraft I

dampers are provided at the nanhole and in the blower discharges to reduce leakage.

The following conditions must exist to permit the leakage of con-taminated air from hot cells:

1.

Failure of utility power, 2.

Failure of energency bus power, 5-1 I

I June, 1986 3.

Failure of standby engine to start.

It is concluded that three such events limit the credibility of such 1

an accident.

5.1.2 Ruptured Fuel Element There is the possibility that a shielded cask falling into the hot I

cell pool might cause a research reactor fuel element to rupture.

The worst possible condition would be the sudden and gross release of all fission gases in the transfer canal. Except for iodine, I

these gaseous fission products would escape to the cask handling roon; however, the cask handling room is maintained at a negative pressure with respect to the outside environment. Exhausted air and any gaseous fission products would pass into the hot cell, through the absolute filters, and up the 50-meter stack.

The point of maximun concentration of a release from a 50-meter I

stack during noderately stable conditions is 3300 meters downwind, as given in Figure A.4, TID-24190.(1)

An individual at this point would receive a naximun dose of 0.0545 Rens as shown in Table 5-1.

The data presented in Table 5-1 were taken from the following I

reference publications:

Column 2 - Table 7.1, Reference 1.

I Column 3 - Tables 5 and 6, Reference 2.

Column 4 - Table 7.1, Reference 1.

Column 5 - Figure A.4, Reference 1.

I I

I I

I i

I I

5-2 I

June, 1986 I

TABLE 5-1 DOSE FROM GASE0US FISSION PRODUCTS I

Colunn 1 Column 2 Column 3 Column 4 Column 5 Column 6 I

Total Ci mci /cc In Fuel Equiv To Max Ground

Dose, Isotope Element (E) Mev 10-_3 Rem /h Conc, mci /cc Rens Kr-85 180 0.24 4.3 0.5 x 10-6 0.0001 Xe-133n 1,500 0.84 1.24 0.42 x 10-5 0.0034 Xe-133 55,000 0.19 5.48 1.52 x 10-4 0.0280 I

Xe-135 14,000 0.62 1,67 0.39 x 10-4 0.0230 Total Dose:

0.0545 I

As an example of the nethod used to calculate the dose from each isotope, I

the calculations for Xenon-133 are presented below. The following as-sumptions were used for these calculations and to establish the values in Table 5-1.

1.

The release occurs over a one-hour period.

2.

The element has been cooling for one day.

3.

The element has operated at 0.5 f1W for one year.

I 4

The iodine is trapped in the water and does not escape.

Reference 2 gives the formula 2.5 x 10-5 0 3

Y=

(Ci/m )

u I

for calculating the maximun concentration. For Xe-133 (Tl/2 = 5.3 days)

I for maximum concertration would be 1.52 x 10-4 pCi/cc, demonstrated as follows:

I 3

.5 x 10-5 55 x 10 3600 15.2 x 10-5 Ci/m 3

i

=

=

2.5 5-3 I

I June, 1986 I

Exposure at the concentration p Ci/cc (from Ref. 2, page 22)

MPC

=

a E(E)

I will give an exposure of 1.0 Ren in one 40-hour week.

I t.

To increase the exposure to 1 Ren, the concentration must be increased by a factor of ten (10).

I 2.6 x 10 p Ci/cc MPC

=

a E(E)

I For the concentration to give a 1 Ren exposure in one hour, the concen-tration nust be increased by an additional factor of forty (40).

-3 1.04 x 10 MPC Ci/cc

=

a E(E)

I The concentration of the cloud radioactive gas that exposes a person standing in the center of it to 1 Ren/h is as follows:

I

-3 1.04 x 10 I

Dose (1 Rem /hr)

Ci/cc

=

E(E) 1.04 x 0-3 5.48 x 10-3 Ci/cc.

1 Rem /hr uCi/cc

=

=

I The exposure at the maxinun concentration is the ratio of the maximun I

concentration (15.2 x 10-5 ci/n3) to the concentration that will give 1 Ren/h (5.48 x 10-3 Ci/cc) or:

1.52 x 10-4 gg 5.48 x 10-3 I

5-4 I

L

I June, 1986 I

I From this analysis, it has been shown that the total dose an individual would receive in an accident described in this section would be 0.054 Rens.

This is an acceptable exposure for an accident.

5.1.3 B&W fiarli B Fuel Assembly Rupture Assumptions 1.

One Mark B fuel assembly is dropped or crushed causing the rupture of all 208 fuel pins.

2.

Fuel assembly burnup, 40,000 MWD /T.

3.

Fuel assembly cooling time,150 days.

I 4

30% of all noble gas escapes.

5.

10% of total iodine is released to the pool water.

6.

The pool decontanination factor for iodine is 100.

7.

All krypton gas released in the pool escapes into the Cask Handling Area.

It then is pulled through the hot cells, absolute filters, and is exhausted up the 150-foot stack.

8.

Release occurs over a two-hour period.

Section 5.12 describes a fuel elenent rupture in the Cask Handling Area.

I In the referenced analysis, the krypton inventory considered was 180 C1.

Ilsing similar assumptions for the nethod and distribution of release, but an inventory for PWR fuel of 6.51 x 103 Ci, the calculated krypton exposure I

at 3300 meters (point of highest ground level concentration do.en wind)

(nearest actual residence to LRC is 1000 iteters) is-I 1 x 10 3

1.1 x 10-3 Rem.

x 6.51 x 10 Ci x 0.3

=

2 1.8 x 10 Ci The iodine release for this accident is calculated using Safety Guide 25 and the above referenced analysis.

F I F PBR (X/Q) 9 D

=

0F DF p

f

-4 3

6 (0.1)(1.39 Ci)(1)(1)(3.47 x 10 m /S)(1.48 x 10 )(2.5 x 10-5)

=

(100)(1)

I 5-5

I June, 1986 Uhere D

0.018 mrad

=

thyroid dose (rads).

D

=

Fg fraction of fuel rod iodine inventory in fuel rod void space

=

( 0.1).

I core iodine inventory at time of accident (curies).

l

=

F fraction of core damaged so as to release void space iodine.

=

fuel peaking factor.

P

=

breathing rate = 3.47 x 10-4 cubic meters per second (i.e.,10 B

=

cubic meters per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> work day as recomended by the ICRP).

DFp effective iodine decontamination factor for pool water = 100.

=

DFf effective iodine decontaninatior factor for filters (if

=

I present) = 1.

j X/0 atmospheric diffusion factor at receptor location (sec/m3),

=

R adult thyroid dose conversion factor for the iodine isotope of

=

interest (rads per curie). Dose conversion are listed in Table I, TID-14844.(3) factors for Iodine 131-135 I

derived fron " p1ndard nan" parameters recomended in ICRP These values were Publication 2, I

5.1.4 Sodium Potassium Fire - Hot Cell An accident in the hot cell could be a fire caused by the ignition I

of the sodium-potassium alloy used in irradiation capsules. Since the use of combustible or flannable materials is severely re-stricted, the area of conflagration would be limited to the capsule I

i tsel f.

Fire extinguishers are available for imediate use through control mechanisms mounted in the cell face and plumbing to the actual incell work area. Fire is not expected to enter the venti-lation system under these coniitions.

The occurrence of explosions is quite unlikely, since explosive materials or gases are not routinely handled. Where solvents are I

used for decontamination, adequate ventilation is provided, and volatile material is linited to quantities that, when vaporized and nixed throughout the volume of the hot cell, would not result in the I

accumulation of an explosive mixture.

l 5-6 I

1

I June, 1986 5.1.5 Zircaloy Fire, Hot Cell 5.1.5.1 General As a part of post-irradiation exanination of PUR spent fuel, the fuel rods are cut into sections with a wetted abrasive cutting blade. The grindings are collected in a water-filled, shallow I

metal pan. The grindings are mixed with " Metal X" fire ex-tinguishing medium and transferred to a 4-inch diameter by 12-inch long sealed radioactive waste container after no more than ten cuts have been made.

An accident is postulated wherein the zircaloy grindings burn in the collection pan in the hot cell.

5.1.5.2 Analysis of Accident It is assumed that multiple operator and supervisor errors occur I

which allow grindings from 100 rod cutting operations to accumu-late in the collection pan, flaterial from 100 cuts would include about 16 grans of zircaloy and 1.6 x 102 grams of spent fuel which I

contains about 2 curies of plutonium.

It is also assumed that the water evaporates from the collection pan so that the grindings become dry. Auto-oxidation of the exposed zircoloy grindings is I

postulated to ignite all zircaloy grindings, thereby releasing 4

4 x 10 calories of heat in a very short period of tim be expected for zircaloy grindings burning in air.(b) e as would It is assumed that the intensity and turbulence of the fire would cause some of the plutoniun-bearing spent fuel to become airborne in the hot cell and that 1% of the plutonium, which was in the I

collection pan, is carried in the off-gas to the HEPA filter. The heat of conbustion is dissipated in the hot cell to the extent that the heat in the off-gas does not damage the filter. The HEPA filter is required to be 99.95% effective.d)

The of f-gas from I

the HEPA filter is released into the stack plune.

5.1.5.3 Results of Accident A total of (2 Ci Pu)(0.01)(0.0005) = 10 pCi Pu is released to the i

environment. The maximum possible amount of Pu in a breathing zone is calculated to be 5 x 10-5 nCi (see calculation below).

I The maximun allowable lung burden for plutonium is 16 nCi. No estimates have been made of the actual amount of plutonium which would be retained in the lungs. Such a consideration would reduce the actual burden roughly an order-of-nagnitude.

I 5-7 I

I June, 1986 5.1.5.4 Conclusion of Accident Analysis The postulated accident would result in a maxinun possible exposure to the public of less than one millionth of a maximum I

allowable lung burden for plutonium.

5.1.5.5 Calculation of Postulated Accidental Dispersion of Plutonium Basis 1.

The nethod presented by Slade(7) may be use.

2.

Release and exposure occurs over a 600 second period.

3.

The breathing rate for an exposed person is 5 x 10-4 m3/sec.

4.

Stack height is 50 meters.

5.

Meteorological conditions are moderately stable (Pasquill F).

6.

Average wind speed in direction of dispersion is 2.5 m/sec.

Fron TID-24190,(8) Figure A.4, the maximum ground level concen-tration is 3300 neters down wind and the dispersion factor 2.5 x 10-5 n-2 Where:

x p/Q'

=

Q' release rate, 17 nCi/sec

=

=

O c

E average wind speed.

=

maximum concentration at ground level.

x

=

-5

-2 (17 nCi/sec) 1.7 x 10 nCi.

-4

2. 5 x 10 g

7 2.5 m/sec 3

m (concentration)(breathing rate)(exposure Amount breathed

=

I period)

I (1.7 x 10-4 nCi)

( 5 x 10-4 *sec ) (600 sec)

=

3 m

5 x 10-5 nCi.

=

I 5-e I

June, 1986 I

5.1.6 2 solute Filter Failure, Hot Cell I

A mechanism for the failure of the absolute filters cannot be postu-lated, but for the sake of analysis, the following assumptions are 4

made:

1 1.

The filters fail.

2.

The radioactive material is released over a 600-second period and is dispersed up the stack.

3 The filters are contaminated with a naximum of one curie of Ru-106 (this assumption is consistent with the fact that the I

filters are unshielded, and one curie of activity is about the maximun that could be allowed without too high a gamma back-ground in the working area).

l I

1 i

The height of the stack is 50 meters (h = 50 neters), and the i

accident is assumed to occur during moderately stable conditions; I

therefore, the maximun concentration is 3300 neters downwind, as given in Figure A.4, TID-24190.(8)

This is off the site; however, it is the point of maxinun ground concentration, and the exposure would be less at other places. Using the formula in Figure A.4, I

TID-24190, the concentration at this point is 1.67 x 10-8 Ci/cc, shown as follows:

2.5 x 10-5 (m-2)

X i/Q'

=

where I

2. 5 x 10-5 g, X

I 3

Q' 1.67 x 10 pCi/s,

=

=

60 I

E 2.5 m/s.

=

2.5 x 10-5

-2 (1.67 x 103 m

Ci/s),

y 2.5 m/s I

1.67 x 10-2 pCi/m3

-8 X

1.67 x 10 Ci/cc.

=

=

(

' I 5-9 I

l June, 1986

\\

The itPC for the general population (as stated in Table II, Colunn 1, 10 CFR 20) is 2 x 10-10 C1/cc for a 168/ hours exposure. Since the iI postulated accident is a 10-minute exposure, an individual would receive 0.083 MPC at the point of maximun concentration. This is demonstrated by the following:

I 1.67 x 10-8 10 min I

2 x 10-10 60 x 168 0.033.

x

=

,g This exposure to an individual (0.083 of the MPC allowed for one 5

week) is acceptable for an accident.

I I

I I

I 1

I l

I I

I lI 5-10 I

I

June, 1986 REFERENCES 1

i I

l 1 Meteorology and Atomic Energy 1968, USGP0, TID-24170, Figure A.4 (1968).

2 " Report of Committee II on Permissible Dose for Internal Radiation,"

I Health Physics, Vol. 3, June 1960.

I 3 Dose Conversion Factors Taken from " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, J. J. DiNunno, R. E. Baker, F. D. Anderson, and R. L. Waterfield (1962).

4 Recommendations of the International Commission on Radiological Pro-tection, " Report of Committee II on Permissible Dose for Internal Radiation (1.959)," ICRP Publication 2 (New York:

Permagon press, 19601 5 Fire Protection Handbook,13 Ed., National Fire Protection Association, 1969, p 5-79.

6 USAEC License SNM-778, Docket 70-824, February 15, 1974, Condition 21, 7 Slade, D. H., Meteorology and Atomic Energy 1968, USAEC, July 1968, p 163.

8 Ibid, p 410.

I-I I

I I

I I

5-11 I

Jun:, 1986 I

6.0 EFFLUENT AND ENVIRONMENTAL MEASUREMENTS 6.1 PRE 0PERATIONAL ENVIRONMENTAL PROGRAMS I

Environmental monitoring prior to construction and operation of the first facility in 1956 was not perfomed.

6.2 OPERATIONAL MONITORING PROGRAMS I

6.2.1 Radiological Monitoring Program 6.2.1.1 Effluent Monitoring Airborne effluents that are potentially contaminated are exhausted through tne 50-neter stack, where practicable. This stack is sampled continuously. Sanple air is drawr. through a fixed filter which is routinely changed and counted on a low background, gas I'

flow proportional counter to determine gross alpha and beta activity. The sensitivity of this counting systen is 8 x 10-17 pCi/nl for gross alpha for the present counting period. Airborne effluents that cannot practicably exhaust through the 50-neter I

stack are individually sampled if there is the potential for these streans to contain 10% or greater of the applicable 10 CFR 20 limits. These samples are counted as described above.

Liquid sampling is perfomed on each of the waste water tanks prior to discharging to the liquid waste treatment systen at the Naval Nuclear Fuel Division. Tanks are stirred and a one-quart I

sample withdrawn. A neasured amount of this sample water is evaporated to dryness on a planchet and counted in a low back-ground, gas flow proportional counter for gross alpha and gross I

beta. The sensitivity of this systen is 3 x 10-7 pCi/n1 for gross alpha and 3 x 10-7 uCi/nl for gross beta for the present counting pe ri od. Ganna spectroscopy is used for isotope identification if the gross technique results in unusually high activities.

6.2.1.2 Environmental Monitoring I

The James River is sampled periodically both upstream and down-stream of the NNFD discharge point (see Figure 2-5).

Sanples are evaporated to dryness on a planchet and counted on a low back-I ground, gas flow proportional counter. Samples are counted to detemine gross dipha and gross beta. The lower limit of detection for this system is 3 pCi/L for gross alpha and 5 pCi/L for gross beta, for the present counting period.

Sanples of Janes River silt and plant life in the vicinity of the l

LRC are periodically taken (see Figure 2-5). These sanples are normally analyzed by an off-site commercial fim, t

6-1 I

June, 1986 Rain water is continuously sampled on site. Measured amounts 'are evaporated to dryness and counted on a low background, gas flow I

proportional counter for gross alpha and gross beta. The lower limit of detection for the system is 3 pCi/L for gross alpha and 5 pCi/L for gross beta, for the present counting period.

6.2.2 Chemical Monitoring The liquid effluents from the LRC that potentially contain harmful I

chemicals are released to the liquid waste treatment system of the Naval Nuclear Fuel Division. That division analyzes effluents to chemical constituants and therefore this is not performed at the LRC.

6.2.3 Meteorological Monitoring I

Wind speed and direction monitors are mounted at the top of the 50-meter stack and at a point about midway up the stack. The infor-mation transmitted from there monitors is recorded on a continuous basis.

Outside air temperature is neasured and recorded continuously at locations on and near the stack at elevations of 50 meters and 3 meters.

I I

I I

I I

I 6-2

-