ML19270F199

Environmental Rept in Support of License SNM-778 Renewal Application.Describes Proposed Action,Site & Facility, Environ Effects of Site Preparation & Plant Const,Environ Effects of Accidents,Effluent & Environ Measurements

ML19270F199
Person / Time
Site:	07000824
Issue date:	12/31/1978
From:	BABCOCK & WILCOX CO.
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ML19270F197	List: ML19270F196 ML19270F198 ML19270F199
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ENVR-781231, NUDOCS 7902010329
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• 11 December, 1978 LYNCHBURG RESEARCH CENTER ENVIRON > ENTAL REPUKT Babcock and Wilcox Co Lynchburg Research Center Post Office box 1260 Lymchburg, Virginia 24505 7902010327'

DECEMBER, 1978 TABLE OF CONTENTS Section Pace 1.0 PROPOSED ACTION.................................... 1-1 2.0 TIIE SITE........................................... 2 3.0 Tile FACILITY....................................... 3-1 4.0 ENVIRONMENTAL EFFECTS OF SITE PREPARATION AND PLANT CONSTRUCTION................................. 4-1 5.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS 5-1 6.0 EFFLUENT AND ENVIRONMENTAL MEASUREMENTS............ 6-1

DECEMBER, 1978 1.0 PROPOSED ACTION This environmental report is submitted by the Babcock and Wilcox Company, Lynchburg Research Center to the U.S. Nuclear Regulatory Com-mission in support of its request for renewal of license SNM-778. This action is taken pursuant to Title 10, Code of Federal Regulations, Part 51.

1.1 BACKGROUND

INFORMATION Babcock and Wilcox, a wholly owned subsidiary of J. Ray McDermott & Company, is a major industrial company which manufactures and markets specially engineered industrial products and materials which help per-form essential tasks for utilities, industries, institutions and govern-ments throughout the world. More than half of the Company's business is in the design, manu-facture and erection of energy systems and components. The balance is in specialty steel tubing, refractories, advanced composites, au*.u-ated machinery, valves and process controls. The production of steam supply systems has traditionally been a 1F~ major part of B&W's business. Today, the Company is one of the leaders in the design and production of both nuclear and fossil power generating equipment for utilities, industry, ships, schools and hospitals. The Re >earch and Development Division provides the operating divi-sions of the Company 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 research centers. One is located in Alliance, Ohio, which is also the division headquarters and the other in Lynchburg, Virginia for which this report is applicable. The Lynchburg Research Center (LRC) was first known as the Critical Experiment Laboratory when it began operation in 1956 as a part of the Atomic Energy Division. In 1957 the AEC issued license CX-1 for the 1-1

DECEMBER, 1978 first privately owned and operated critical experiment facility in the United States, which was located at the Laboratory. This facility was used to design and test the first nuclear core for the Consolidated Edison Power reactor. This was a thorium core, the first of its kind built in the U.S. In 1958 additions to the Critical Experiment Laboratory included facilities for the nuclear merchant ship critical experiment, the Lynch-burg Source Reactor (CX-12), and the Lynchburg Pool Reactor (R-47). The Laboratory expanded again in 1964 with the addition of the Nuclear Fuels Laboratory. This building included the Babcock and Wilcox Test Reactor (BAWTR), an oxide fuel preparation laboratory, and a hot cell 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 Laboratorj was added. This facility was built to accomodate the equipment necessary to process plutonium mixed oxide fuel preparation and examination. The LRC presently employees 232 scientists, engineers, technicians -y_ and support personnel. Approximately 50% of the work is performed under NRC licenses. The remainder is in t'.e areas of process control, non-destructive examination methods and instrument development, and non-nuclear ceramics. Research and development utilizing source, by-product and special nuclear material is performed in three buildings primarily. Building A houses one Critical Experiment facility (CX-10) and a one megawatt pool type research reactor, the Lynchburg Pool Reactor (LPR), R-47. The CX-10. is a tank type facility licensed for a maximum of 1 KW power utilizing low enriched uranium dioxide fuel. It is used for physics verification experiments, computer code verifications and benchmark t es tin g. The LPR is licensed for a maximum power of 1 MW utilizing g, highly enriched uranium. This reactor is used for reactor operator 1-2

DECE3BER, 1978 training, neutron radiography, neutron transmission measurements, activation analysis, resonance integral measurements, instrument development, irradiation of experime.nts in the associated autoclave, reactivity measurements and source preparation. Building B houses the hot cell facility, the crane and cask handling area, a radicchemistry laboratory, a counting laboratory, and a scanning electron microscope laboratory. The four hot cells are used to handle and examine materials that are highly radioactive. Ir-radiated commercial nuclear fuel assemblies have been partially dis-assembled and destructive and nondestructive examinations performed on the fuel rods. Reactor irradiated experiment capsules have been disassembled and studied and examinations of primary system components are performed. The cask handling area, the radiochemistry lab and the scanning electron microscope lab, support the hot cell operations. Research and Development of new and improved methods of nuclear fuel preparation is performed in Building C. This building is equipped with glove boxes for handling alpha emmitting isotopes, standard chem-ical fume hoods, a perchloric acid fume hood, X ray em=ission and defrac-tion equipment and an emmission spectrograph. This building was con-structed for testing methods of plutonium mixed oxide fuel production. im Presently such testing is limited to using benchscale quantities of plutonium. Scrap recovery studies utilizing depleted uranium are being performed and a program to study the reprocessing of thorium based. fuels by chloride volatility is underway. This latter program utilizes thorium, depleted uranium, and the nonradioactive species of fission product elements. The Radioactive Waste Storage Building is used to house contain-erized radioactive solid was' e subsequent to shipping for off-site disposal. The Liquid Waste Disposal Facility is a " tank f. arm" where process area liquid wastes are collected, stored, sampled, diluted, and pumped to the waste disposal facility of the Naval Nuclear Fuels Division. 1-3

DECEMBER, 1978 1.2 REGIONAT SITE LOCATION The selection of the Mt. Athos site as the location of the major portion of B&W's nuclear fuel activities was based on a number of criteria related to social, environmental and economic factors. The validity of the choice has been demonstrated by the successful conduct of nuclear-related activities at the site since 1956, 1.2.1 Social Factors The Lynchburg area possesses an exceptionally stable and productive work force. Additionally, the acceptance of nuclear activities as a safe and valuable industry has resulted in excellent community relations with B&W. The company is not aware of any adverse reaction to the location of the facility on the outskirts of Lynchburg. Lynchburg, which is one of the major industrial centers in Virginia, has a solid and varied industrial base, a pleasant climate, active community organizations, and plentiful recreational opportunities. 1.2.2 Ecomonic Factors The LRC provides its services principally for the company's operating iF~ divisions. Three of its customers, the Nuclear Materials Division, the Naval Nuclear Fuels Division, and the Nuclear Power Generation Division are located in the Lynchburg area. Other company facilities extensively utilizing LRC services are located in Gcorgia, 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 commercial 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 economic penalty asscciated with a large labor turnover. 1-4

DECEHBER, 1978 1.2.3 Environmental Considerations The LRC is located approximately six miles from the nearest major 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 determined its acceptability for nuclear activities on the basis of geological stability, groundwater flow characteristics, and hydrological considerations relating to the neighboring Jam ts River. The maximum flood crest recorded for the James River was approximately 100 feet below the LRC. Additionally, the Center is only minimally affected by storms moving 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 therefore this section is not applicable. 1.4 PkEVIOUS ACTION ON APPLICATION Table 1.1 provides a detailed history of AEC and NRC licensing activities relating to the LRC. Tables 1.2 and 1.3 give the same infor- "r-mation for the Commercial Nuclear Fuels Plant and the Naval Nuclear Fuels Division respectively, which share the same 525 acre site. 1-5

DECEMBER, 1978 TABLE 1.1 AEC AND NRC LICENSE ACTIVITIES FOR THE LYNCHEURG RESEARCH CENTER Date Activity March, 1957 CX-1 Issued February, 1958 CX-10 issued September, 1958 CX-12 isaued September, 1958 R-47 issued May, 1962 CX-19 issued February, 1964 TR-4 issued March, 1964 SNM-778 issued September, 1966 SNM-778 reissued (incorporating licenses 45-105-3, SMB-714, SNM-32, and SNM-744 c February, 1972 CX-12 terminated March, 1973 TR-4 terminated June, 1973 CX-1 terminated June, 1973 CX-19 terminated March, 1974 SNM-778 renewed 1-6

DECEMBER, 1978 TABLE 1.2 AEC AND NRC LICENSING ACTIVITIES FOR THE COMMERCIAL NUCLEAR FUELS PLANT Date Activity February, 1969 SUB-969 issued December, 1969 SIC 3-1168 issued June, 1970 SNM-1168 incorporates SUB-969. SUB-969 terminated. May, 1975 Amendment to SNM-1168 authorizing pellet fabrication approved. July, 1976 SNM-1168 renewal 9 1-7

DECEMBER, 1978 TABLE 1.3 AEC AND NRC LICENSING ACTIVITIES FOR THE NAVAL NUCLEAR FUELS DIVISION Date Activ ity August, 1956 SNM-42 issued August, 1961 SNM-42 extended June, 1965 SNM-42 renewed March, 1977 SNM-42 renewed r 1-8

DECEMBER, 1978 FIGURE 1.1 BABCOCK AND WILCOX COMPANY ORGANIZATION CHART Board of Directors President Research & Development Division Vice President -a Lynchburg Research Center Alliance Research Center Director Director 1-9

DECEHBER, 1978 FIGURE 1.2 LYNCHBURG RESEARCH CENTER ORGANIZATION CHART Lynchburg Research Center Director Accounting & Purchasing Personnel Administrative Facilities Services g r Materials and Chemical Advanced Controls and Technology Experimental Physics Laboratory Laboratory 1-10

DECEMBER, 1978 2.0 THE SITE 2.1 SITE LOCATION AND LAYOUT The Lynchburg Research Center is located on the James River about 4 miles east of Lynchburg, Virginia. The site, which comprises 525 acres (not all of which is used for the LRC), lies within Campbell County and borders on Amherst County, Figure 2.1 shows the location of the LRC in relation to major population centers within the state. Figure 2.2 shows the population centers and physical features within 5 miles of the LRC. The irregularly shaped property is bounded on three sides by a large loop of the James River and on the remaining side by State Route 726, which closely follows the base of Mount Athos. This mountain rises rapidly from about 500 feet MSL to 900 feet MSL, making it the dominant feature of the surrounding landscape. The Babcock and Wilcox property consists of large sections of relatively flat floodplain along the James River lying at about 470 feet MSL. The interior of the property is largely composed of rolling hills, one of which rises to almost 700 feet MSL. Figure 2.3 shows the property beundary and depicts the topography within about a two mile radius of the LRC. The land in the immediate vicinity of the plant is sparsely in-habitated. The severe topography makes it unsuitable for commercial farming and the boundaries of nearby Lynchburg have not yet pushed this far out into the country. The Lynchburg Foundry, a producer of light metal castings, occupies a parcel of land which abuts the south boundary of the Babcock and Wilcox property. The Foundry is about 1/2 mile from the LRC. The site is serviced by a spur of the Chesapeake and Ohio Rail-road which runs through the Babcock and Wilcox property. The property is also conveniently located fer truck and automobile access. About three miles from the LRC State Route 726 connects with U.S. Highway 460, a major link between Roanoke and Richmond. The City of Lynchburg is serviced by the Preston Glenn Airport from which ten flights leave daily f or Washington, D.C., and other southern, eastern, and midwestern terminals. 2-1

DECEMBER, 1978 of the 525 acres owned by 3&W at this location only 13.6 acres are utilized by the LRC. Other major B&W tenants at the site are the Naval Nuclear Fuels Division and the Commercial Nuclear Fuels Plant. The acreage assigned to each of these "acilities is given in Table 2.1 and the locations of these separate facilities within the site e indicated in Figure 2.4. A 0.7 acre privately-owned cemetary lies within the B&W property bounds. This area is not affected by normal operation or maintenance of the facility. Access is granted to the owner by easement. 2.2 REGIONAL DEMOGRAPHY, LAND AND WATER USES Information applicable to the Lynchburg Research Center is found in: Environmental Report Babcock and Wilcox Commercial Nuclear Fuel Plant Lynchburg, Virginia December, 1974 License SNM-1168, docket 70-1201 2.3 REGIONAL HISTORIC,SCEN-IC, CULTL'RAL tWD NATURAL LANDMARKS Information applicable to the Lynchburg Research Center is found in: Environmental Report Babcock and Wilcox Commercial Nuclear Fuel Plant Lynchburg, Virginia December, 1974 License SNM-1168, docket 70-1201 2.4 GEOLOGY Information applicable to the Lynchburg Research Center is found Ln: Environmental Report Babcock and Wilcox Commercial Nuclear Fuels Plant Lynchburg, Virginia December, 1974 License SNM-ll68, docket 70-1201 2-2

DECEMBER, 1978 2.5 HYDROLOGY Information applicable to the Lynchburg Research Center is found in: Environmental Report Babcock & Wilcox Commercial Nuclear Fuels Plant Lynchburg, Virginia December, 1974 License SNM-1168, docket 70-1201 2.6 METEOROLOGY Information applicable to the Lynchburg Research Center is found in: Environmental Report Babcock & Wilcox Co=sercial Nuclear Fuels Plant Lynchburg, Virginia December, 1974 License SKH-1168, docket 70-1201 2.7 ECOLOGY Information applicable to the Lynchburg R2scarch Center is found

s. _

in: Environmental Report Babcock & Wilcox Commercial Nuclear Fuels Plant Lynchburg, Virginia December, 1974 License SKH-1168, docket 70-1201

2.8 BACKGROUND

CHARACTERISTICS Regional radiological data has been collected by the LRC since 1963. Samples of vegetation, river silt, river water and air are routinely col-lected by the LRC. Analyses are performed on site when possible or by an outside contractor when equinment or techniques are not locally available. Lata collected for the three previous years (1975, 76, 77) are presented in tables 2.2 thru 2.8. Figure 2.5 shows the general locations of sample poin t s. 2-3

DECEMBER, 1978 TABLE 2.1 ALLOCATION OF LAND AT THE MT. ATHOS SITE Acres Lynchburg Research Center Security fenced land occupied by buildings 4.0 (building ground area - 2.25 acres) Land occupied by parking areas and driveways 2.01 Additional land suitabic for future building 7.6 (tentatively allocated) Subtotal 13.6 Commercial Nuclear Fuel Plant Security fenced land occapied by buildings 2.5 (building ground area - 1.5 acres) Land occupied by parking areas and driveways 2.6 concidered for future building Additional land suitable for future building 20.0 (tentatively allocated) Subtotal 25.1 Naval Nuclear Fuel Division Security fenced land occupied by buildings 19.0 (building ground area - 9.5 acres) Land occupied by parking areas and driveways 17.3 considered for future building Additional land suitable for future building 199.0 Land unsuitable for building 251.0 Su5 total 486.3 Total 525.0 2-4

DECEMBER, 1978 TABLE 2.2 VEGETATION 3 AMPLE RESULTS GROSS ALPHA (10~ u Ci/ gram) Year Quarter East of B&W West of B&W 1975 1 1.7 + 1.2 1.9 + 1.3 2 0.0 + 1.0 2.1 + 0.7 3 0.0 + 1.0 2.1 + 0.9 4 1.6[1.4 1.4[1.3 1976 1 11.6 + 5.0 9.7 + 4.7 2 6.2[2.9 5.1[2.8 3 2.8 1 1.8 3.0 1 1.9 4 1.0 1 0.6 1.2 1 0.7 1977 1 4.9 + 3.0 7.3 + 3.0 2 2.9 + 1.2 2.3 + 1.1 3 1.5 + 1.0 2.2 + 1.5 4 3.4 1 2.0 3.9 1 2.0 Gross Beta - K-40 1975 1 11.3 + 1.2 14.2 + 1.4 4_. 2 6.9 lI O.3 11.9 II 0.6 3 6.1 II 0.4 8.6 II 0.6 4 6.5 1 1.1 9.3[1.2 1976 1 5.7 + 0.8 3.5 1 0.7 2 20.6 1 1.7 21.1 + 1.8 3 26.0 + 2.0 7.0 + 0.5 4 10.0 1 1.0 23.0 1 2.0 1977 1 15.0 + 2.0 15.0 + 2.0 2 7.6 1 1.0 14.0[1.0 3 17.0 + 2.0 8.4 + 2.0 4 22.0 1 3.0 19.0 + 2.0 A 2-5

DECEMBER, 1978 TABLE 2.3 JAMES RIVER SILT SAMPLE RESULTS GROSS BETA (10" u C1/ gram) 1/2 Mile 1/2 Mile 2 Miles 5 Miles Year Quarter Above B&W Below B&W Below B&W Below B&W 1975 1 14.2 1 1.2 9.5 1 1.0 9.2 1 1.0 6.9 1 0.9 2 3.3 + 0.7 4.2 + 0.8

4. 2 + 0.8 1.7 + 0.5 3

2.0 I 0.4 4.1 I 0.6 4.9 I 0.6 1.1 I 0.3 4 10.6 1 1.4 6.4[1.2 7.0 1 1.2 6.5[1.2 ~ Annual Average 6.0 1 0.9 1976 1 1.6 + 0.6 5.7 + 1.1 0.9 + 0.5 3.1 + 0.9 2 8.1 I 0.9 8.4 I 0.9 13.3 I 1.3 8.1 I 0.9 3 0.0 I 0.5 0.8 I 0.2 7.0 I 0.8 0.0 I 0.5 4 13.0 1 2.0 17.0 1 2.0 10.0[2.0 16.0 1 2.0 Annual Average 7.1 1 0.3 1977 1 0.7 + 0.1 2.4 + 0.3 0.0 + 0.5 0.0 + 0.5 2 5.8 I 0.6 5.0 I 0.6 5.3 I 0.6 5.8 + 0.6 3 8.8 I 1.4 0.0 I 0.5 0.0 I 0.5 0.0 I 0.5 4 1.4[1.0 4.2[0.8 8.4[1.0 0.0[0.5 Annual Average 3.8 1 0.2 2-6

DECEMBER, 1978 TABLE 2.4 JAMES RIVER SILT SAMPLE RESULTS GROSS ALPHA (10~ p C1/ gram) 1/2 Mile 1/2 Mile 2 Miles 5 Miles Year Quarter Up Stream Below B&W Below B&W Below B&W 1975 1 4.1 1 3.1 6.4 1 2.8 5.2 1 2.6 6.6 1 2.8 2 2.911.9 3.1 1 1.9 4.1 1 2.1 3.4 1 2.0 3 2.9 1 1.4 3.5 1 2.2 4.6 1 1.7 1.6 + 1.2 4 3.9 1 1.8 3.0 1 1.7 2.9 1 1.7 3.3 t 1.8 Annual Average 4.2 1 2.0 1976 1 0.0 1 1.0 10.8 1 4.9 0.0 1 1.0 0.0 1 1.0 2 2.9 1 2.4 6.4 1 3.0 4.5 1 2.7 6.0 1 2.9 3 0.0 1 1.0 0.0 1 1.0 8.5 1 5.5 5.8 1 3.5 4 3.4 1 2.1 3.5 1 2.0 3.0 1 2.0 3.0 1 2.0 Annual Average 3.6 1 0.7 m. 1977 1 3.2 1 2.7 5.9 1 3 6.413 4.7 1 3 2 11 17 13 +7 4.5 1 3.7 5.0 1 3.8 3 6.6 + 5.6 7.1 + 5.7 0.0 + 1 0.0 + 1 4 0.0 1 1 0.0 1 1 0.0 1 1 0.0 1 1 Annual Average 4.2 1 0.9 2-7

DECEMBER, 1978 TABLE 2.5 JAMES RIVER WATER SAMPLE RESULTS ~ GROSS ALPHA (10 ' u C1/ml) ~ 1975 1976 1977 Up Down Up Down Up Down Month Stream Stream Stream Stream Stream Stream Jan. 3 3 3 3 3 3 Feb. 3 3 3 3 3 3 Mar. 3 3 3 3 3 3 Apr. 3 3 3 3 3 3 May 3 3 3 3 3 3 Jun. 3 3 3 3 3 3 Jul. 3 3 3 3 3 3 Aug. 3 3 40 3 3 3 Sep. 3 3 3 3 3 3 Oct. 3 3 3 3 3.8 3 Nov. 3 3 3 3 3 3 Dec. 3 3 3 3 3 3 Note: Limit of sensitivity = 3 x 10-d- 4 C1/mi 2-8

DECEMBER, 1978 TABLE 2.6 JAMES RIVER WATER SAMPLE RESULTS GROSS BETA (10- U C1/ml) 1975 1976 1977 Up Down Up Down Up Down Month Stream Stream Stream Stream Stream Stream Jan. 3 3 3 3 3 3 Feb. 3 3 8.9 6.4 8.6 6.4 Mar. 6.1 3 3 3 5.7 3 Apr. 4.4 4.4 3 4.7 3 3 May 3 4.5 3 3 5.7 5.3 Jun. 3.2 3.7 3 3 4.3 3 Jul. 9 7.5 3 3 3.4 5 Aug. 4.5 3 4.2 3.4 8.2 9.3 Sep. 4 3 9.4 16.5 11.0 7.0 Oct. 3 13 3 5 10 12 Nov. 3 3 3 3 3 3 Dec. 24 3 4 3 4.5 4 Note: Limit of sensitivity = 3 x 10- p C1/ml 4_. 2-9

DECEMBER, 1978 TABLE 2.7 MONTHLY AVERAGE BACKGROUND AIR CONCENTRATIONS Beta Alpha -14 (10 p C1/ml) (10-p Ci/ml) Month 1975 1976 1977 1975 1976 1977 Jan. 5.0 1.4 8.0 2.1 1.5 1.8 Feb. 7.9 1.4 2.4 1.4 1.9 0.9 Mar. 9.2 1.5 6.5 0.6 0.7 1.48 Apr. 13.8 1.63 9.8 0.6 0.9 2.1 May 7.7 1.44 18.9 0.6 0.7 1.2 Jun. 7.5 1.2 18.0 1.9 1.06 0.86 Jul. 4.3 1.52 15.5 1.6 1.6 0 89 Aug. 3.7 1.17 8.9 3.8 2.2 0.71 Sep. 1.4 1.35 15.9 0.66 0.94 3.4 Oct. 3.9 13.0 21.0 1.1 0.90 1.5 Nov. 1.3 12.4 9.0 1.2 1.05 1.39 Dec. 0.95 6.2 4.1 1.8 1.04 1.29 Note: Air samples ar.1 taken on site, east of Building A. c 2-10

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DECEMBER, 1978 3.0 THE FACILITY 3.1 EXTEP'ul APPEAPECE The buildings that comprise the LRC are all of masonry construction. 3.1.1 Buildine A Building A is constructed of concrete block basically. The walls of the critical experiment bays are poured concrete. That portion of the building which f aces the Naval Nuclear Fuels 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 multipanc, horizontal rectangles. 3.1.2 Buildine 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 contain first and second floor vertical rectangular windows accented at the top and bottom by green stone slabs. The seventh contains a second floor vindow and the front door. The Build-ing is 340 feet by 98 feet. The south lawn is landscaped with evergreen scrubs and hemlock trees. The remainder of the building is surrounded by a lawn of grass. 3.1.3 Buildine C Building C is a single story building of concrete block construction. 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 con-tains the eight windows of the building, and its front door. The block face is covered with painted stucco. A driveway abutts the front left-hand portion of the front of the building and the rear right-hand por-tion. The front right-hand portion is a grassy lawn with evergreen scrub landscapin g. Those areas of the building net abutted by the driveway are grassy lawns. 3-1

DECEMBER, 1978 3.1.4 Building J Building J is the solid waste storage facility. It is located in the rear of Euilding C. This building is a single floor concrete block square structure. The building contains no windows. A 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 north face. The building exterior is painted a beige. The building is surrounded by asphalt paving and a chain link fence. 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. Thic buildinb 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 abutt: it on the east. 3,1.6 Building D Building D is a complex of six buildings. Five of these are single floor, concrete block buildings with grey agrigate brick facing on all sides. The central building is two stories high 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 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 OPEPATION Operations at the LRC are widely diverse and change frequently. A brief description of current operations is given in the sections that

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

3-2

DECDiEER, 1978 TABLE 3.1 1977 USACE Water 364000 gal./ month Natural Gas 932000 ft.3/ month Electricity 353000 KWH / month TABLE 3.2 EFFLUE!!T Air (Total Volume) 1.104 x 10 ft /yr Liquid Waste (released to NNFD 419,500 gallons treatment system 1977) 6 Liquid Waste (released to NNFD 2.s44 x 10 gallens sanitary sewage system, 1977) Liquid released to the James 1.82 x 10 gallons River (1977) Solid nonradioactive waste (1977) 2.25 x 10 ft 3 Solid radioactive waste (1978) 1.3 x 10 ft 3-3

DECEMBER, 1978 TABLE 3.3 LIOUID WASTE RELEASES TO NNFD TPR T"ENT SYSTEM -6 (10 Curies) 1975 1976 1976 1977 1977 1978 JUL-DEC JAN-JUN JUL-DEC JAN-JUN JUL-DEC JAN-JUN 1/. 8 Cr-51 21.0 253.3 18.1 5.6 7.6 Mn-54 82.5 1043.0 26.8 6.2 2.7 Co-58 Co-57 0.19 Co-60 95.5 82.0 1111 99.7 72.1 151.5 Fe-59 7.6 Zn-65 Sr-90 20.25 8.7 80.4 13.3 42.1 8.6 Y -90 20.25 8.7 71.3 13.3 42.1 8.6 Nb-95, Zn95 8 0.3 12 0.7 Ru-106 28.4 Cs-134 407.3 67.5 25.4 5.7 58.5 96 Cs-137 5138,8 1067 348.6 73.3 1146 1990 Ba-140, La-140 Ce-144 2.3 Gross beta 82.6 15.0 159.1 260.2 72.1 248 Uranium 44.3 15.6 57 37.2 188 80.5 3-4

DECEMBER, 1978 TABLE 3.4 LRC AIRBORNE RELEASES FROM 50 ME~ER STACK 1977 January thru June Activity Concentration ~ Gross alpha particulate < 0.04 pCi < 6.7 x 10 pCi/mi Gross beta particulate <2 pCi < 3.3 x 10-10 pCi/ml -1 Kr-85 <10 milli ci < 2.1 x 10 pCi/ml July thru December Activity Concentration -19 Gross alpha particulate < 0.053 pCi < 1.13 x 10 pC1/ml -18 Gross beta particulate < 1.5 pCi < 3.2 x 10 C1/ml -11 Kr-85 < 8.7 Ci < 1.86 x 10 pCi/ml -12 E-3 < 0.71 Ci < 1.52 x 10 pCi/ml ~1978 January thru June ' Activity Concentration -0 Gross alpha particulate < 0.016 pCi < 3.8 x 10 C1/ml -1 Gross beta particulate < 0.97 pCi < 2.3 x 10 pCi/ml Kr-85 < 0.1 Ci < 2.4 x 10 ~3 pC1/mi E-3 < 0.01 Ci < 2.4 x 10" pCi/c1 Note: In the above table, actual concentrations of long-lived particulate activities in the stack discharge are less than background activities, because a large portion of the air in the stack enters through "abso-lute" filters. Since the concentration is less than background there is no reasonable way to measure what contributions come from environ-mental background or from laboratory processes. Therefore, the value presented is a maximum and not only includes environmental background but in all probability is environmental background. 3-5

DECEMBER, 1976 TABLE 3.5 SOLID RADIOACTIVE k'ASTE SHIPPED FOR LURIAL (1978) Total Volume = 1.3 x 10 ft Material Amount Co-60 579 C1 Cs-137 0.2 Ci Mixed Fission Products 467.8 Ci H-3 2 Ci U-235 72 Grams Fu 35 Grams U - depleted 155 Kilograms 3-6

DECEMBER, 1976 3.2.1 Building A - There are primarily four programs underway in Building A at the present time. 3.2.1.1 L3mehburn Pool Reactor (LPR) The LPR operates pursuant to NRC license R-47, docket 50-99. This facility is utilized for instrumentation development, neutron radiography, operator training, neutron ac :;ation analysis and neutron transmission studies. Tne effluents given below appear also in the facility's annual report for 1977. The LPR has no defined gaseous exhaust stream (Fig. 3-1); therefore, tne quantities listed here are estimated. Gaseous: ~ Total noble 6 activation gases 2 x 10 Ci Normal steady state operations 1.3 x 10~ pCi/ml ~ Maximum instantaneous 3 x 10 pCi/ml Liquid: Total gross beta 7.01 x 10~ Ci Average beta concentration 1.8 x 10~ pCi/ml -6 Total alpha < 2.2 x 10 Ci Average alpha concentration < 2.6 x 10~ pCi/ml Total vol. of liq id discharged 3.8 x 10 t 3.2.1.2 Critical Exoccirent Facility (CX-10) The CX-10 facility operates pursuant to NRC license CX-10, docket 50-13. This facility presently is used in the conduct of a DOE sponsored study of close packed fuel storage. The operations of this facility does not produce detectable gaseous releases (figure 3-1). Liquid releases are very small and are combined with those of the LPR for analysis. Nonradio-active solid waste is included in the total for the Center. 3-7

DECEMBER, 1978 3.2.1.3 Computer Studies Several projects fall into this category. All generate no radioactive waste and nonradioactive solid waste is included in the totals for the Center. 3.2.1.4 Ceramics A project is underway in that area of Building A designated as Bay 1. The purpose of the prcject is to study the effects of heat and pressure on vessels lined with a protective ceramic lining. This study produces solid, nonradioactive waste which is disposed of in onsite land fill. Approxi-3 mately 600 ft per year is disposed of in this f ashion. The material is primarily aluminum oxide. 3.2.2 Building B This facility is co= prised of four hot cells, a hot cell operations area, a cask handling area, a transfer canal and storage pool, a small instrument repair shop fer work on contaminated equipment, an experimental pool, a nuclear and radiochemistry laboratory, two metallurgy laboratories, t nting laboratory, a health physics counting area, a ceramics oven room, a machine shop, a scanning electron microscopy lab and a fracture mechanics lab. Radioactive solid, liquid and gaseous releases are combined in the totals for the Center, as is the nonradioactive effluents. Figures 3-2, 3-4 show the facility ventilation and liquid waste systems. 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 instrument repair shop. The transfer canal and storage pool is used to receive, unload, 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 recirculated through ion exchange columns 3-8

DECEMBER, 1978 for cleanup. These resins are replaced when expended and handled as dry waste. Particulates that settle to the pool bottoc are removed periodically with an underwater vacuum cleaner and disposed of as dry waste. The hot cells are used to perform destructive and nondestructive testing and examination of highly radioactive materials. These include reactor core hardware components and fuel rods removed from irradiated reactor fuel assem-blies. The cells generate solid and gaseous radioactive wastes. The gaseous wastes consist of krypton which originates from fuel rods that are punctured for fission gas analysis. The iodine component is decayed prior to shipment to the LRC. High level solid wastes are placed in special containers, removed from the cells and placed in below grade storage tubes to await shipment in shielded shipping containers off site. The instrument repair shop is used when repair of manipulators is required and for performing work on items that are radioactive but not to the extent that remote hot cell handling is required. Solid radioactive vaste is generated in the area. The cask handling area is a high bay room used to receive and ship containers of radioactive material. The largest source of waste is generated in decontaminating shipping containers. Liquid waste in the form 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 with the non-destructive analyzers. No radioactive vastes are generated in this area. Nonradioactive solid waste is included in the totals for the LRC. 3.2.2.2 Experimental Pool This 30,000 gallon pool is used to develop underwater examination equipment. Radioactive material is not new handled in this pool. 3.2.2.3 Nuclear and Radiochemistry Laboratory This laboratory utilizes standard chemical fume hoods the exhausts of which pass through one prefilter and one HEPA filter. One exception to this 3-9

DECEMBER, 1978 is a perchloric acid fume hood which exhausts directly to the roof of Euilding B. At present, no work utilizing radioactive material is per-formed in the latter hood. Work of interest being presently performed is analysis of irradiated fuel samples, corrosion products, neutron flux dosimeters and reactor coolant samples. Low level radioactive wastes are released through the liquid waste dispcsal facility. Other liquid wastes are solidified for of' site burial. Solid waste is shipped for off site burial. Airborne and gase...s effluents are filtered and discharged through the 50 meter exhaust stack. All these contributions are included in the site totals. 3.2.2.4 Metallurry Laboratory The metallurgy laboratory has equipment for structural examinations on a macrescopic and microscopic scale. Facilities are available for all metallography preparations and examinations utilizing light-microscopy. A hot stage metallograph is available for microscopic examination of mate-rials at high temperatures and in controlled atmospheres. An industrial x-ray unit is also available to this laboratory. Wastes from the metallurgy laboratory are typically nonradioactive and solid. Water used for cooling is discharged to the storm drains. 3.2.2.5 countine Laboratory The counting laboratory contains several high resolution gamma spectro-scopy systems coupled to computere for data processing. A liquid scintil-lation system is used for spectroscopy of low energy beta emitters. Gross counting and spectroscopy are performed on alpha and beta emitting elements. The laboratory is not equipped with sample preparation facilities. Preparation is performed in other labs and transferred to the counting lab and returned after counting to the sending lab. No releases are made from this laboratory. 3-10

DECEMBER, 1978 3.2.2.6 Ceramics Oven Room This room is used for mixing, forming and sintering nonradioactive ceramic materials. Wastes are primarily solids that are included in the LRC solid waste totals. Cooling water is discharged to the storm sewer. 3.2.2.7 Scanning Electron Microscopv Laboratory Radioactive end nonradioactive specimens are prepared and examined in this facility. Small amounts of solid wastes are generated and these are included in the LRC totals. 3.2.2.8 Fracture Mechanics Area This area contains a closed-loop electrohydraulic load frame, an impact tester and a fatigue precracker. Specimens are brought into this laboratory for testing and returned to the originating lab for disposal. 3.2.3 Building C Building C provides 20,000 square feet of laboratory, office, and support space. The building was originally designed for handling multi-kilogram quantities of plutonium. The present license limit severely restricts the amount of plutonium that can be handled but the facilities for handling large quantities remain. The research and development per-formed in Building C primarily involves the use of unirradiated source and special nuclear materials. Some work involving the use of by-product material is carried out in the facility but it is very limited. The building off gas system is shown in figure 3-3. The liquid waste drain system is shown in figure 3-5. Radioactive solid waste is packaged for off site burial and transferred to the Solid Waste Storage Building to await shipment. Nonradioactive waste solid is disposed of with other such vaste in municipal land fill. 3-11

i e DECEMBER, 1978 3.2.3.1 Analvtical Chemistry Analytical chemistry is performed in laboratories 19, 20, and 27 (figure 3-6). The following major ite=s of equipment are available for this use: 1. I-ray diffraction 2. Emission spectrometer 3. Ato:ic absorption spectrometer 4. Polarograph 5. Gas chromatograph 6. Spectrophotometers 7. Carbon, sulfur, and halide analyzers 8. Moisture analyzer 9. Differential thermal analyzer 10. Subsieve nizer Additionally there are numerous items of equip =ent needed to do traditienal wat chemicci analyses. Nonradioactive liquid trastes are diluted and released to the liquid waste disposal systes. Liquid wastes containing radioactivity are evaporated to dryness, solidified, or disposed of through the liquid waste disposal system, depending on the amount of radioactive caterial. 3.2.3.2 Process Develourent Process develepcent is performed in laboratories 43, 44 cnd 50 (figure 3-6). This work is in the following areas of the nuclear fuel cycle: I 1. Fuel conversion 2. Fuel, control and coderator caterials fabrication 3. Scrap recovery 4. Effluent treatment 5. Warte treatment (i. Nonaqueous, nongaseous coolants Liquio wastes are handled as in 3.2.3.1. 3-12

DECEMBER, 1978 3.2.3.3 Fuel Materials Development Fuel Materials Development work is performed in laboratories 15, 16 and 17 (figure 3-6). The follov'ing major items of pilot-plant-scale equipment are available for this work. 1. Thirty ton hydraulic press 2. High temperature, hydrogen atmosphere pusher furnace 3. High temperature, hydrogen atmosphere periodic furnace 4. High temperature, cold wall furnace 5. Centerless grinder 6. Miscellaneous powder blenders 7. Bench metallograph and associated ceramography equipment 8. Glove box line containing complete pellet and vipac fabrication lines Liquid radioactive waste is not generated in these areas. 3.3 WASTE C0hTINEMENT AND EFFLUD;T CONTROL 3.3.1 Air Effluents The exhaust air from the LRC is made up of two streams, air exhausted from hoods, glove boxes, hot cells, and potentially contaminated areas, and general building air necessary to maintain comfort. Exhaust air from hoods, glove boxes and hot cells are passed through a prefilter and at least one stage of HEPA filtration prlcr to release via the 50 meter high stack. Room off gas from area where there exists the potential for airborne radioactive contamination is passed through c prefilter and one stage of HEPA filters and is released through vents at essentiaJ1y roof height. General building air is partially recirculated for energy conserva-tion and released at roof height. 3-13

DECDEER, 1978 3.3.1.1 Controlled Area Air Effluents Exhausts from hot cells, fume hoods and glove boxes are the main sources of supply to the 50 meter high stack. This stack is sampled isokinet-ically continually while work in these areas is in progress. Drawings of the systems are shown in figures 3-1, 3-2, and 3-3. t.ir passing into the stack has been filtered through at least one stage of HIFA filters. In the case of tha hot cells, glove boxes and Building C fume hoods, two series stages are used. Two perchloric acid fume hoods presently listalled, are exceptions to the above practice. These hoods exhaust directly to the roof of Buildings B and C with no filtration. Room off ges from labs in Puilding C is passed through one stage of HEPA filters and exhausted through a 5 meter stack of which approximately 50% is recirculated. Releases through the 50 meter stack are given in table 3-4. 3.3.1.2 Nonradioactive 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 1977. On the basis that IJO percent of the =aterial evaporated and was released through the ventilation system 1.82 lb/ day would result. 3.3.2 Liquid Effluents Liquid effluents leava the LRC by three routes, the storm sewer which not only carries rain water but the major portion of cooling water, the sanitary sewage line which flows to the treatment facility at the Naval Nuclear Fuels Division (NNFD) and the only noteworthy one of the three, the effluent from the liquid waste retention tanks which flows into the treatment facility of NNFD. 3.3.2.1 Contaminated Licuid Waste System Potentially contaminated and contaminated liquid wastes from labor-atories are directed to the liquid waste disposal system. The drain sys-tecs for Buildings B and C are shown in figures 3-4 and 3-5. A schematic 3-14

DECEMBER, 1976 diagram of the liquid waste retention tanks and piping is shown in figure 3-7 Enilding A wastes, from the two reactors, drain tc a below grade 5000 gallon tank located outside the northeast wall of the building. All waste tanks are thoroughly mixed and sampled prior to release to the NNFD system. Effluents must meet the limits of release to an unrestrit-ted area, given in 10 CFR 20, prior to release. If sampling indicates that the tank contents exceed this restriction, dilution is used. A compilation of releases through this system is given in table 3-3. 3.3.2.2. Sanitarv Waste Effluents Untreated sanitary wastes are combined for treatment with those of the NNFD's at that f acility's sanitary waste treatment facility. The LRC's 3 contribution to this facility is 5.6 x 10 8allons per day. 3.3.2.3 Storm Drainace Runoff from the parking lot, building roofs and surrounding land exits the LRC on the east side of the site, and flows through a natural dry stream bed to the James River. Water used for furnace cooling and similar uses drain into this system at a rate of 5000 gallons per day. The quality of this water is the same as the site process water. 3.3.3 Solid Wastes All solid wastes generated from LRC operations are monitored and disposed of as described below. Ceramic pressure vessel liners are an exception to this. These liners are not radioactive and are disposed of in a land fill on site at a rate of 600 ft per year. 3.3.3.1 Contaminated Solid Wastes Contaminated solid wastes are disposed of by a NRC licensed facility. These wastes consist of filters, packing material, decontamination equipment, contaminated laboratory equipment and solidified liquids. These wastes are packaged and stored at the LRC until a sufficient amount has accumulated for shipment to burial. Packaged wastes are 3-15

DECEMBER, 1978 stored in a building specified for this purpose. A fenced area adjacent to this building is used for storage of pachaged LSA and fissile exempt material. 3.3.3.2 Uncontaminated solid wastes Approximately 2.25 r. 10 cubic feet of uncontaminated solid wastes is generated at the LRC per year. These wastes are routinely monitored to ensure that they are not radiologically contacinated and disposed of by a private contractor at the Lynchburg sanitary landfill. Salvageable materials, such as metals, are sold or recycled, i l 1 3-16

FIGURE 3-1 A ( 1 PA ROOF e g jr (1200) pA A PA PA ' MAIN VENTILATION DUST O ITOO 3 ta d'2000 J'1200 l' . t.(1000) M (450LO PA ---i H f \\) D D tf t One Way Damper tn j, -h - Damper h t T CHN ,, g y o,,r - Filter Box [ BH -Bolonce Hood e PH -Perchloric Hood 0) I SH - Storage H -Furne Hood of -LDING 'C' P - Pr e-Filt er PA -Pre-Filter 8 Absolute F!!ter () h -Pressure Indicator 25000 hl @ -Pressu*e Indiestor 150 F00T ST ACK l- -Cop

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DECEMBER, 1978 4.0 ENVIRONMENTAL EFFECTS OF SITE FREPARATION AND PIANT CONSTRUCTION AND OPEPATION 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 land-scaped or allowed to return to its natural state. Any social, economic, or ecolcgical impacts of construction are now history. Undoubtedly construction aided the economy of the region by providing employment, and presumably small areas of land were removed from biological productivity and dedicated to research and development use. There is no evidence to indicate that the social, economic, or eco-logical impacts of construction were harmful, or even of a very great magnitude. 4.1.1 Land Use The site was originally used primarily for farming and consequently much of the land had already been cleared long before construcrion 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 vegetation before construction have been reforested with pine trees and grassy mea-dows. The natural landscape has been altered to accommodate buildings, parking lots and access roads. Each of these has been designed to mini-mize undesirable environmental effects. Overall, these alterations have not had an adverse effect on terrestrial life. Significant 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 construction of parking lots, roadways, and buildings did not significantly alter the natural drainage patterns of surface water flow. 4-1

DECEMBER, 1978 4.2 EFFECTS OF PLANT OPERATIONS 4.2.1 Radiological Iceact 4.2.1.1 Airborne Effluents As stated in the note in table 3.3, air releases from operations at the LRC are in all probability attributable entirely to background for ling lived particulate activity. During irradiated fuel examina-- tions Kr-85 is released. Tabic 3.3 shows that about 10 Ci of Rr-85 may be released as a result of these operations in a year. The exposure to a person living in the City of Lynchburg from this type of release is 0.00005 man-rens assuming the following: The variability of the winds to Lynchburg which e ie due west of the LRC is 4.5%. The City of Lynchburg is five miles west of the e LRC. The population of Lynchburg is 60,000 people, e e An individual exposed to an integrated cloud of 1 Ci sec/m = an exposure of 7 x 10 ' Rem. 3 4.2.1.2 Liquid Effluents Referring to figure 3.3 during the period of July thru December, 1975, 5.1388 mil 11 curies of Cs-137 was released to the NNFD waste treat-ment system. This is the highest release of the period covered. The man-rem exposure for a release of five mil 11 curies of CS-137 per year to the James River is given below.

1. The nearest municipal water cystem utilizing the James River, down stream of the LRC is the city of Richmond.

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

4-2

DECEM3ER, 1978 3. The average flow rate of the James River is 3 5500 ft per second. e Concentration at Richmond: 5 x 103 u C1 4 ml/ft3) (5500 ft3/sec)(365 days)(24 hr/ day)(3600 sec/hr)(2.83 x 10 = 1.02 x 10~ p Ci/ml e An adult drinks 370 liters of water per year The dose ingestion conversion factor is e -5 7.14 x 10 cre=/pCi ingested e Activity ingested: ~ (1,02 x 10~ p C1/ml)(3.70 x 10 ml) = 3.8 x 10 p Ci Man-rem exposure in Richmond is: e -5 (0.38 pC1)(7.14 x 10 mrem /pci) (250,000 people) = 6.7 man-millirem Release records for years previous to those reported in Table 3.3 show this release to be somewhat above the average six month release periods. 4.2.2 Chemical Discharge Chemical discharges from the LRC are made through the liquid waste disposal system. Based on the receipts of hazardous chemicals less than a liter per day is discharged. These discharges are made to toe NNFD liquid waste treatment system and is included in that facility's sample results. 4.3 RESOURCES COMMITTED The following resources were committed for the facility: 1. The land 2. Structural materials 4-3

DECEMBER, 1978 3. Nonrecoverable consumables used during construction. The only land permanently affected by construction was the four acres 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 the dedication of 6.1 acres fcr the facility has no noticeable effect on the natural ecosystem, and it does not permanently foreclose other options for human development. 4.4 DECOMMISSIONING AND DISMANTLING

Reference:

J Babcock & Wilcox Res.. arch & Development Division Lynchburg Research Center Decommissioning Plans License SNM-0778 April 18, 1978 4-4

DECEMBER, 1978 5.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS 5.1 CENERAL 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. 5.1.1 Power Failure, Hot Cell 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 develops. In this assumed situation, the Hot Cell ventilation is maintained by one fan connected to the emergency bus. One fan is adequate to maintain a AP of 0.25 inch of water over the cell face. However, emergency power from the motor generator is pro-vided to both fans, and this emergency power source is checked once a week to ensure startup. This motor generator is equipped with an auto-matic startitig mechanism and a backup manual starter if the automatic starter should fail. Hot cell emergency lighting is sufficient to per-mit limited operations to safely secure the cell. Ventilation is maintained through the normal duct work, which contains a prefilter and absolute filters that remove particulate materials. The Hot Cell operations that produce radioactive gases are handled in such a way that these gases are contained. Fission gases can only be released to the cell atmosphere by manual operation of a valve. The gas is released only after an estimate of gross activity is complete. Gas release of this type is allowed only during normal ventilation conditions and is stopped immediately in the event of an emergency. Thus, it is apparent that, even with a complete loss of power to the facility, the surrounding area is adequately protected. The hot cell ventilation air joins that from the clean areas in the manhole behind 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 5-1

DECEMBEh, 1976 system, since the suction from the stack fan would provide an air velocity greater than 100 fpm from the manhole. Backdraft dampers are provided at the manhole and in the blower discharges to reduce leakage. The following conditions must exist to permit the leakage of contam-innted air from hot cells: 1. Failure of utility power, 2. Failure of emergency bus power, 3. Failure of standby engine to start. ~ lt is concluded that three such events limit the credibility of such an accident. 5.1.2 Ruptured Fuel Element There is the possibility that a shielded cask falling into the hot The cell pool might cause a research reactor fuel element to rupture. worst possible condition would be the sudden and gross release of all fission gases in the transfer canal. Except for iodine, these gaseous the fission products would escape to the cask handling room; however, 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 maximum concentration of a release from a 50-meter stack during moderately stable conditions is 3300 meters downwind, as given in Figure A.4, TID-24190. (1) An individual at this point would receive a maximum dose of 0.0545 Rems as shown in Table 5.1. The data presented in Table 5.1 were taken from the following ref erence publica-tions: Column 2 - Table 7.1, reference 1. Column 3 - Tables 5 and 6, reference 2. Column 4 - Table 7.1, reference 1. Column 5 - Figure A.4, reference 1. 5-2

DECEMBER, 1978 i i i TABLE 5.1 DOSE FROM CASEOUS FISSION PRODUCTS Column 1 Column 2 Column 3 Colu=n 4 Column 5 Column 6 Total Ci mci /cc in fuel equiv to Max ground

Dose, Isotope element I(Z) MeV 10-3 Rem /h cone, mci /cc rems

_6 85 Kr 180 0.24 4.3 0.5 x 10 0.0001 133 5 Xe 1,500 0.84 1.24 0.42 x 10 0.0034 m 133 -4 Xe 55,000 0.19 5.48 1.52 x 10 0.0280 135 _4 Xe 14,000 0.62 1.67 0.39 x 10 0.0230 Total dose: 0.0545 As an example of the method used to calculate the dose from each isotope, the calculations for xenon-133 are presented below. The fol-lowing assumptions 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 MW for one year. 4. The iodine is trapped in the water and does not escape. Reference 2 gives the formula -5 2.5 x 10 q, 3 p, C1/m ) U 133 for ca..:ulating the maximum concentrantion. For Xe (T = 5.3 days) /2 the maximum concentration would be 1.52 x 10 uCi/cc, demonstrated as follows: 3 -5 55 x 10 -5 3 2.5 x 10 3600 = 15.2 x 10 C1/m. E = 2.5 a 5-3

DECDBER, 19 78 Exposure at the concentration -6 MPC),= 2.6 x 10 C (fr C Page 22, Reference 2) I(E) will give an exposure of 0.1 Rem in one 40-hour week. ~ To increase the exposure to 1 Rem, the concentration must be increased by a factor of ten (10). pCi/cc MPC)a = I(E) For the concentration to give a 1 Rem exposure in one hour, the concen-tration murt be increased by an additional f actor of forty (40). MPC),= pCi/cc The concentration of the cloud radioactive gas that exposes a person standing in the center of it to 1 Rem /h is as follows: Dose (1 Rem /h) = u cc IE -3 1 Rem /h pC1/cc = 5.48 x 10 pCi/ce. = 9 The exposure at the maximum concentration is the ratio of the maximum -5 concentration (15.2 x 10 ci/m ) to the concentration that will give -3 1 Rem /h (5.48 x 10 pCi/cc) or: 1.52 x 10' 0.028 Rem. = -3 5.48 x 10 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 Rems. This is an acceptable exposure for an accident. 5.1.3 B&W Mark B Fuel Assembly Rupture Assumptions 1. One Mark B fuel assembly is dropped or crushed causing the rupture of all 208 fuel pins. 5-4

DECDGER, 1978 2. Fuel assembly burnup, 40,000 E'D/T. 3. Fuel assembly cooling time, 150 days. 4. 30% of all noble gas escapes. 5. 10% of total iodine is released to the pool water. 6. The pool decontamination factor for iodine is 100. 7. All krypton gas released in the pool escapes into the Cask Ilandling Area. It then is pulled through the hot cells, absciute filters, and exhausted up the 150 foot stack. 8. Release eccurs over a two hour period. Section 5.12 describes a fuel element rupture in the Cask Ilandling Area. In the referenced analysis, the krypton inventory considered was 180 C1. Using similar assumptions for the method and distribution of 3 release, but an inventory for PWR fuel of 6.51 x 10 C1, the calculated krypton exposure at 3300 meters (point of highest ground 1cvel concen-tration down wind) is: 3 -3 x 6.51 x 10 Ci x 0.3 - 1.1 x 10 Rem 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 PER (X/Q) E D= DF DF p f , (0.1)(1.39 Ci)(1)(1)(3.47 x 10- m /s)(1.48 x 10 )(2.5 x 10-5) 6 (100) (1) = 1.78 x 10-rad or 0.018 millirad. Where D = thyroid dose (rads). F = fraction of fuel rod iodine inventory in fuel rod void space (0,1). I = core iodine inventory at time of accident (curries). F = fraction of core damaged so as to release void space iodine. F = fuel peaking factor. 5-5

DECDBER, 1978 B = breathing rate = 3.47 x 10 cubic meters per second (i.e., 10 cubic meters per 8 hour work day as recommended by the ICRP). DF = effective iodine decontaminati n factor for pool water. DF = effective iodine decontamination factor for filters (if present). f X/Q = atmospheric diffusion factor at receptor location (sec/m ). R = adult thyroid dose conversion factor for the iodine isotope of interest (rads per curie). Dose conversion factors for Iodine 131-135 are listed in Table 1. TID-14844.(3)* These values were derived from " standard man" parameters recommended in ICRP Publication 2 (4) 5.1.4 Sodium-Potassium Fire - Hot Cell An accident in the hot cell could be a fire caused by tha ignition of the sedium-potassium alloy used in irradiation capsules. Since the use of combustible or flammable materials is severly restricted, the area of conflagration would be limited to the capsule itself. Fire extinguishers are available for immediate use through control mechanisms mounted in the cell face and plumbing to the actual incell work area. Fire is not expected to enter the ventilation system under these condi-tions. The occurrence of explosions is quite unlikely, since explosive materials or gases are not routinely handled. ktere solvents are used for decontamination, adequate ventilation is provided, and volatile material is limited to quantities that, when vaporized and mixed throughout the volume of the hot cell, would not result in the accumulation of an explosive mixture. 5.1.5 Zircaloy Fire, not Cell 5.1.5.1 General As a part of post-irradiation examination of Pk'R spent fuel, the fuel rods are cut into sections with a vetted abrasive cutting blade. The grindings are collected in a water-filled, shallow metal pan. The grindings are mixed with " Metal X" fire extinguishing medium and trans-ferred to a 4-inch diameter by 12-inch-long sealed radioactive waste 5-6

DECEMBER, 1978 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 which allow grindings frcm 100 rod cutting operations to accumulate in the collection pan. Material from 100 cuts would include about 16 grams of zircaloy and 1.6 x 10 grams of spent fuel which contains about 2 curies of plutonium. It is also assumed that the water evaporates from the collec-tion pan so that the grindings become dry. Auto-oxidation of the exposed s zircaloy grindings is postulated to ignite all zircaloy grindings, thereby releasing 4 x 10 calordec of heat in a very short period of time as would be expected for zircaloy grindings burning in air. It is assumed that the intensity and turbulence of the fire would cause some of the plutonium-bearing spent fuel to become airborne in the hot cell and that 1% of the plutonium, which was in the collection pan, is carried in the off-gas to the HEPA filter. The heat of combustion is dissipated in the hot cell to the extent that the heat in the off-gas does The HEPA filter is required to be 99.95% cffective.(6) not damage the filter. The off-gas from the HEPA filter is released into the stack plume. 5.1.5.3 Results of Accident A total of (2 Ci Pu)(0.01)(0.0005) = 10 pCi Pu is released to the environment. The maximum possible amount of Pu in a breathing zone is -5 calculated to be 5 x 10 nCi (see calculation below). The maximum allowable lung burden for plutonium is 16 nC1. 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 toughly an order-of-magnitude. 5.1.5.4 Conclusion of Accident Analysis The postulated accident would result in a maximum possible exposure to the public of less than one millionth of a maximum allowable lung 5-7

DECDEER, 1978 burden for plutonium. 5.1.5.5 Calculation of Postulated Accidental Discersion of Plutonium Basis The method presented by Slade( ) may be used. 1. 2. Release and exposure occurs over a 600 second period. -4 3 3. The breathing rate for an exposed person is 5 x 10 m /sec. 4. Stack height is 50 meters. 5. Metc.orological conditions are moderately stable (Pasquill F). 6. Average wind speed in direction of dispersion is 2.5 m/sec. From TID-24190,(0) Figure A.4, the maximum ground level concentration is 3300 meters down wind and the dispersion factor E s/Q' = 2.5 x 10~ ~ m Where: Q' = release rate, = n sec. Ose 5 = average wind speed. x = maximum concentration at ground level. -5 -2 (17 nci/sec), 1.7 x 10 nci , 2.5 x 10 m 2.5 m/see m Amount breathed = (concentration)(breathing rate)(exposure period) (* ~ (5 x 10 )(600 sec) = m -5 = 5 x 10 nci. 5.1.6 Absolute Filter Failure, Hot Cell A mechanism for the failure of the absolute filters cannot be postulated, but for the sake of analysis, the following assumptions are made: 1. The filters fail. 2. The radioactive material is released over a 600-second period and is dispersed up the stack. 5-8

DECEMBER, 1978 3. The filters are contaminated with a maximum of one curie of 106Ru (this assumption is consistent with the fact that the filters are unshielded, and one curie of activity is about the maximum that could be allowed without too high a gamma background in the working area). The height of the stack is 50 meters (h = 50 meters), and the accident is assumed to occur during moderately stable conditions; therefore, the maximum concentration is 3300 meters downwind, as given in Figure A.4, TID-24190.( ) This is off the site; however, it is the point of maximum ground concentration, and the exposure would be less at other places. Using the formula in Figure A.4, TID-24190, the concentration at this point is 1.67 x 10- pCi/cc, shown as follows: i p/Q' = 2.5 x 10 g-2) -5 where -5 7, 2.5 x 10 0' y 3 Q' = 1 Ci = 1.67 x 10 pCi/s, 600s p = 2.5 m/s. -5 3 7, 2.5 x 10 -2 (1.67 x 10 uCi/s), 2.5 m/s -2 -8 i = 1.67 x 10 C1/m = 1.67 x 10 pCi/cc. The MPC for the general population (as stated in Table II, Column 1, -10 10 CFR 20) is 2 x 10 Ci/cc for a 168/ hours exposure. Since the postu-lated accident is a 10-minute exposure, an individual would receive 0.083 MPC at the point of maximum concentration. This is demonstrated by the following: 1.67 x 10 min = 0.083. 2 x 10 60 x 168 This exposure to an individual (0.083 of the MPC allowed for one week) is acceptable for an accident. 5-9

DECEMBER, 1978 5.1.7 Fire and Explosion in Buildinc C Building C is a steel-framed, cinder block building especially designed to contain plutonium. Its metal ceilings are sealed to prevent the spread of contamination outside of the building. Special coatings have been applied to the floor to facilitate decontamination. The building offgas and air-conditioning systems are designed to preclude the release of plutonium as long as there is no breach of the building's integrity. Although a reasonable credible accident condition that would release plutonium to the environment cannot be postulated, we have assumed that a release has occurred in the following manner: a. A fire starts by spontaneous combustion of loose trash in a glove box under moderation control. (Loose trash is a violation of operating procedures). b. Next the building off-gas fails so that room air can seep out of the building. This seepage would take place until remedial action was taken. Ascuming the above then: 1. 0.01% of glove-box plutonium escapes to the room and is airborne. 2. 10% of the contaminated room air escapes from the building. 3. Exposure and release occurs over a 600-second period, -4 3 4. The breathing rate for an exposed person is 5 x 10 m /s, 5. The distance.to the site boundary is 500 meters, 6. The glove box contains 2000 grams of plutonium ( Pu 91%, Pu 8%, Pu 0.8%) with a specific activity of 13.4mg/mC1. The release rate could then be calculated as follows: 2000 g x 0.0001 x 0.1 - 1.5 mci or 1500 pCi are released during a 10-minute period; therefore, 5-10

DECEMBER, 1978 1.5 mci 2.5 pCi/s is the release rate. = 600 s The concentration of plutonium at the site boundary can be calcula-ted as follows using equations 3.141 and 3.142 from TID-24190: 2 7, Q' eY h WIZ p 25 25 Y Z Ey = y + (CA/x) oy' + (CA/s){ 1/2 Z = By substituting the values below in the formula, we arrive at the concen-tration rate at the site boundary (1.5 x 10~ pC1/m ). 5= time average concentration at a point (Ci/m ), Q'= release rate (Ci/s), height of release (m) = 0, h = mean wind velocity in the x-direction (m/s) 2 2. y = dispersing coefficient 2 0.5, C = y y displacement (m) = 0, = oy lateral diffusion for moderately stable conditions = 20 = (Figure 3.10, reference 4), vertical diffusion for moderately stable conditions = 9 az = (Figure 3.11, reference 4), building cross-sectional area = 275 m, A = 0 0 2I + 2I i 2.5 pCi/s e v z 1.5 x 10 ' pCi/m. ~ = n x 48 x 44 x 2.5 -4 3 With a breathing rate of 5 x 10 m /s and a 600-second exposure period, an individual's breathing zone would contain 0.45 x 10~ pCi plutonium, demonstrated by the following: 1.5 x 10~ pCi/m x 5 x 10 ' x 600 = 0.45 x 10 ' pCi plutonium. ~ ~ 5-11

DECEMBER, 1978 The maximum lung burden for plutonium is 0.016 uC1. The breathing zone would contain less than 1/100 of a lung burden. No estimates are included for the amount of plutonium retained in the lungs over the amount inhaled. Such a consideration would reduce the concentration by approximately a factor of ten. 5.1.8 Criticality Accidents 5.1.8.1 General LA-3611 A review of thirty-five criticality incidents (the thirty-four incidents listed in LA3611 and the 1972 Windscale incident) show seven have occurred in a production chemical processing environment and the remainder occurring in the course of measuring reactor parameters or running an experimental reactor facility. The processing facilities have had accidents due to unanticipated accumulations of fissile material through breakdowns or deviations from the normally established transfer paths. The criticality accidents have generally occurred during transient periods such as start-up operation (i.e. Y-12 at Oak Ridge and Wood River Junction), or clean-up operations (Hanford Works, April 7, 1962). The order of magnitude of these proces-6 sing facility accidents have been about lx10 fissions on Lne initial 18 spike and lx10 fissions total yield, numbers representative of the Y-12 accident. It is highly unlikely that such a criticality event could happen at LRC since LRC doesn't perform continuous processing or handle the amount of fissile material characteristics of the above pro-duction processing facilities. 17 Historically, high yield (greater than 2x10 fissions) critical accidents have originated from large critical assemblies where the active material is measured in tcres of tons (i.e., EBR-1, SL-1, NR reactor). Probably the most destructive criticality accident was the SL-1 reactor 18 incident. Its total yield and fission spike was abo st 5x10 fissions. The largest known accident (in terms of total fission yield) was the NRX reactor which had 1.2x10 fissions. In ccatrast, the R&D activities at LRC and quantities of SNM handled (doubly batch safe) would seem to preclude such a large critical incident at LRC. 5-12

DECEMBER, 1978 In summary, there has been no known criticality accident associated with the type of R&D activities conducted at LRC in concert with SNM License 778. Using this review as a basis, it would assume to be reasonable that in the assumptions used for evaluating the potential radiological conse-quences of an accidental nuclear criticality at LRC, the assumptionc need not be as severe as those for a fuel fabrication plant (i.e., Regulatory Guide 3.34). Hence, it is assumed that a criticality accident of lx10 total fission yield with a lx10 fission spike be used as the basis for evaluating the potential radiological consequences. A criticality accident could occur anywhere in our facility where SNM Jc handled, although the probability is very low. However from the standpaint or now it effects the environment, the accident must be ana-lyzed in two different cases. 5.1.8.2 Criticality Accident in Buildine A A criticality accident is assumed to occur in Building A using the assumptions ( } in the Regulatory Guide 3.34 along with the following: 1. Building A is not pressure tight. However it is of masonry and concrete construction with doors and windows closed except for personnel access. Therefore the leak rate would be low occurring over several hours. Short-lived fission gases and halogens would undergo decay before reaching off site environment. 2. 100% of the noble gas and 25% of the halogens produced are assumed to escape. 3. Total fissions are lx10 spread over an eight hour period. 4. The closest site boundary is 280 meters, however only material in a radioactive cloud could reach someone at the 280 meter boundary because the facility is located on a plateau and direct radiation would have to go through many feet of earth in order to get there. The closest direct line of sight point is approximately 400 meters. 5-13

DECEMBER, 1978 -3 3 (10) 5. X/Q with wake correction at 280 meters is 1.8x10 sec/m. 6. No plutonium is handled in Building A so no plutonium is involved. The external whole body dose from direct radiation at 400 meters is gamma 0.034 Rems, neutrons 0.055 Rems or a total of 0.089 Rems. The gamma whole body exposure from a semiinfinite cloud of radioactive gas for 280 meters is given in Table 5.2 and is 0.081 Rems. At 400 meters it would be approximately 0.081 x.62 =.050 Rems. Therefore, the total body exposure at 400 meters which is the highest whole body exposure location is 0.088 + 0.05 = 0.138 Rems. Thyroid exposure from radioactive iodine is given 1:1 Table 5.3 and is 1.44 Rems. 5-14

DECEMBER, 1978 TABLE 5.2 DOSE FROM SEMIINFINITE CLOUD OF NOBLE GAS BUILDING A GROUND LEVEL RELEASE AT SITE BOUNDAR'l (280 METERS) Ci Released ( ) Rads /Ci( Dose Rems T 1/2 Isotope 1.F6 hrs Kr-83m 3.7 1.9 x 10 <.000 y 4.4 hrs Kr-85m 1.7 x 10 1.2 x 10 0.001 1.6x10f 1.72 x 10[5 <.000 10 y Kr-85 76 min Kr-87 1.0x10{ 6.17 x 10_3 0.006 2.8 hrs Kr-88 6.6 x 10 1.52 x 10_^ 0.020 3 3.2 min Kr-89 4.1 x 10 1.73 x 10 ~ 0.027 12 d Xe-131m 3.9x10[2 1.56x10[f <.000 2.3 d Xe-133m 5.5 x 10 3.3 x 10_3 <.000 5.3 d Xe-133 1.3 3.5 x 10 0.002 y -3 16 min Xe-135m 1.1 x 10 3.4 x 10 <.000 7 -3 9.2 hrs Xe-135 1.7 x 10 1.9 x 10,3 0.002 3 4.2 min Xe-137 3.9 x 10 1.51 x 10, 0.003 17 min Xc-138 1.1 x 10 9.2 x 10~ 0.020 0.081 Rems Whole Body (1). Regulatory Guide 3.34 " Assumptions used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant", April, 1977. (2). Rads /Ci released 2 (Ci Released)(Factor Ref. 3)(3.17)(X/Q) corrected for decay before it reaches site boundary. Regulatory Guide 1.109 " Calculations of Annual Doses to Man f roni Routine Release of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I", Oct., 1977, Rev. 1 pp. 1.109-21. X/Q - 1.8 x 10-sec/m Regulatory Guide 3.34 pp. 3.34-10 and 11. 5-15

DECDIBER, 1978 TABLE 5.3 IODINE EXPOSURE EUILDING A ACCIDENT AT 280 METERS X/Q x Breathing Pate Isotopes Ci 6.25 x 107 T 1/2 Rad /Ci(1) Rems ( -11 -17 I-129 4.3 x 10 2.7 x 10 1.6 x 10 yr 2.6 x 10 <.000 I-131 0.18 1.12 x 10~ 8.1 d 1.44 x 10 0.162 I-132 0.67 4.18 x 10~ 2.3 hrs 1.50 x 10 0.006 -6 I-133 3.5 2.186 x 10 21 hrs 3.1 x 10 0.676 -5 1-134 4.8 x 10 3.0 x 10 53 min 3.3 x 10 0.100 1 -6 1-135 1.2 x 10 7.50 x 10 6.7 hrs 6.6 x 10 0.495 1.44 Rems (1). Chemical Rubber Company " Handbook of Radioactive Nuclides" (1969) pp. 222. (2). Calculated from equation D = 74fTE/m ibid pp. 204. Report of ICRP Committee II " Permissible Dose for Internal Radiation" (1959). (3). R = (X/Q sec/m )(Breathing Rate m /sec)(Release C1)(Dose Conversion Factor R/C1). 5-16

DECEMBER, 1978 5.1.8.3 Criticality Accident in Buildinc C A criticality accident is assumed to occur in Building C using the assumptions in the Regulatory Guide 3.34.( ) 1. Building C is under negative pressure with respect to the out-side environment. There are two exhaust systems. One is room air (ROG) and the other is hoods and glove boxes (BOG). Both of these systems are HEPA filtered. The ROG system is used primarily to recirculate and filter air to supply heating or cooling as the season demands. The main source of exhaust is the BOG system as it exhausts through two sets of HEPA filters for discharge up the 150 foot stack. The calculation for this criticality accident will assume that the stack is the discharge point for the noble gases, halogens, etc. 2. 100% of the noble gas and 25% of the halogens produced are assumed to escape. 3. Total fissions are lx10 spread over an eight hour period. 4. The closest site boundary is 500 meters. sec/m,(10) 3 -5 5. X/Q for the 50 meter stack is 6x10 6. Plutonium is handled so that case will be considered. 7. If for some reason the discharge is not through the stack the conditions for Release from Building A would apply and the results would be less severe than the Building A case as the distance to offsite is 500 meters and not 280 meters. The external whole body dose from direct radiation at 500 meters is gamma 0.015 Rems and neutrons 0.021 Rems for a total of 0.036 Rems. The gamma whole body exposure from a semiinfinite cloud of radio-active gas for 500 meters is given in Table 5.4 and is 0.005 Rems. Therefore the total whole body exposure at 500 meters is 0.005 +.036 = 0.041 Rems. 5-17

DECDiBER, 1978 TABLE 5.4 DOSE FROM SD11 INFINITE CLOUD OF NOBLE GAS BUILDING "C" 150 FT. STACK RELEASE. SITE BOUNDARY AT 500 METERS C T/12 Isotope Ci Released (1) Rec. Guide 1.109(2) Rems (3) -5 1.86 hrs Kr-83m 3.7 1.9 x 10 <.000 x10[3 <.000 4.4 hrs Kr-85m 17 1.2 _4 10 yrs Kr-85 1.6 x 10 1.72 x 10 <.000 -3 76 min Kr-87 100 6.17 x 10 0.001 -2 2.8 hrs Kr-88 66 1.52 x 10 0.002 -2 3.2 min Kr-89 4100 1.73 x 10 0.001 12 d Xe-131m 3.9 x 10 ' 1.56x10j ~ <0.000 2 2.3 d Xe-133m 5.5 x 10 3.3 x 10 0.000 -4 5.3 d Xe-133 1.3 3.5 x 10_3 <0.000 16 min Xe-135m 11 3.4 x 10- <0.000 0.000 x 10,3 9.2 hrs Xe-135 17 1.9 3 1.51 x 10,3 4.2 min Xe-137 3.9 x 10 <0.000 3 3 17 min Xe-138 1.1 x 10 9.2 x 10 0.001 0.005 Rems (1). Regulatory Guide 3.34 " Assumptions used for Evaluating the Potential Radiological Consequences of Accident of Nuclear Critic lity in a Uranium Fuel Fabrication Plant" (April, 1977). (2). Regulatory Guide 1.109 "Calculatinn of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1". pp. 1.109-21. (3). Rems = (Factor Regulatory Guide 1.109) (31. 7) (X/Q) (Decay Fac tor) 5-18

DECEMBER, 1978 Thyroid exposure from radioactive iodine is given in Tabic 5.5 and is 0.047 Rems. Lung exposure from plutonium released is calculated in Table 5.6 and is 2.26 x 10-6 Rems. TABLE 5.5 IODINE EXPOSURE FOR STACK RELEASE (50 M STACK) Times X/Q x B. Rate Isotope C1 2.08 x 10-8 Rads /Ci(1) Rems (3) -11 -19 7 I-126 4.3 x 10 8.95 x 10 2.6 x 10 <0.000 -9 0 I-131 0.18 3.75 x 10 1.44 x 10 0.005 I-12? 0.67 1.39 x 10 1.5 x 10 <0.000 5 I-133 3.5 7.3 x 10-3.1 x 10 0.022 -8 3 1-134 48.0 1.0 x 10 3.3 x 10 0.003 -6 4 I-135 12.0 2.5 x 10 6.6 x 10 0.016 0.047 Rems (1) Ibid, pp 222 (2) From Chemical Rubber Company " Handbook of Radioactive Nuclides" (1969) Equation D = 74fTE/m, pp 204. Report of ICRP Committee II Permissible Dose for Internal Radiation (1959). (3) Rems = (X/Q)(Breathing Rate)(Rads /C1)(Ci). 5-19

DECEMBER, 1978 TABLE 5.6 DOSE BUILDING C PLUTONIU11 RELEASE T11RCUGli IIEPA FILTERS AND UP 150 FT STACl;

Release, Times Factor (1 Dose Conversion Ci 1.04 x 10-11 Factor (2)

Rems -15 8 -6 Pu-238 5.9 x 10 6.14 x 10 3.16 x 10 1.9 x 10 -5 -16 8 Pu-238 2.7 x 10 2.81 x 10 3.16 x 10 8.9 x 10- -16 Pu-240 5.8 x 10~ 6.0 x 10 3.16 x 10 1.9 x 10- -2 -1 -9 Pu-241 1.8 x 10 1.87 x 10 1.5 x 10' 2.6 x 10 -18 0 -9 Pu-242 4.7 x 10~ 4.9 x 10 3.16 x 10 1.5 x 10 -5 16 8 -8 Am-241 2.4 x 10 2.5 x if 3.16 x 10 7.9 x 10 -6 2.26 x 10 Rems (1) Factor = (X/Q)(Breathing Rate)(Release through IIEPA Filters) 3 = 6 x 10-5 (sec/m3) x 3.47 x 10-4 m /sec x 0.0005 = 1.04 x 10-11 (2) Dose = 74fTE/m Chemicc1 Rubber Company " Handbook of Radioactive Nuclides" (1969), pp 204. 74 (0.15)(500)(57) f = 0.15* 1,000 T = 500 days * = 316 Rems /pCi inhaled 0 s D = 3.16 x 108 Rens/Ci inhaled Dose = Rems E Pu-241 = 0.053*

• Report of ICRP Committec II Permissible Dose for Internal Radiation (1959).

5-20

DECEMBER, 1978 5.1.9 LPR Postulated Accident As stated in Reference (11), the maximum credible accident for the LPR is the rapid addition of an outside fuel element when the reactor is just below or at a delayed critical condition. This could only result through a gross set of human errors and the complete > ;sregard of all operating procedures. A 0.017 reactivity addition would result to cause an initial power surge terminated by a reactor period scram, or by a void formation in the core. If the void formation terminated the first surge, the reactor would then be shut down by period, high power, high radiation, or a manual scram. This maximum credible reactivity accident would result in no core meltdown or contained fission product release. The hazards associated with a reactor operation at 1,000 kW arise chiefly from the production and accu =ulation of fission products. Since the maximum credible accident would not result in the release of significant fission products, a hypothetical accident is assumed. The safety rods are assumed to fail to scram following the maximum credible accident. The reactor is then shut down by melting enough fuel from a central element to compensate for the excess reactivity. This amount of fuel is about one-third of a fuel element. This represents about 2.5% of the total fission products in the core. Assumptions used in calculation: 1. One percent of halogens are carried to top of water along with bubbles of noble gas. Forty percent of noble gas escapes to room air. 2. No steam reaches the top of the water. 3. Less than 1% of solid fission products are carried to the top of the water along with noble gas. 4. The LPR wing is not pressure tight. However, during operation and normally during shut down, all outside doors and windows are kept closed. 5. The reactor has been at 1,000 kW for 1 year. 6. Only 10% of solid fission products escape the building. All noble gas and halogens that reach the room air escape. 7. The release would be over a period of time (at least 8 hrs). Some short-lived radioisotopes would decay significantly before they reached the site boundary. 5-21

DECEMBER, 1978 Table 5.7 shows direct exposure and thyroid exposure from the radioactive cloud passing over an observer at the site boundary (280 meters). The calculations are conservative since: 1. The assumed time of operation is much longer than the reactor could or will run. 2. The probability of winds blowing in the directior. of the minimum exclusion area is very low. 3. The population density outside the exclusion area is very low. 5-22

l l DCCCMBER, 1978 TABLE 5.7 Wi!OLT EODY DOSE A':D TlIYROID DOSE AT EITE DOUNDARY (280 METCi$) TROM !!YPOTliETICM LPI: ACCIDENT 5 Released to Bldg Air Ci x 0.00025 Rems Isotope Corc{l) T 1/2 C1 Rads /Ci( ) (8 hr exposure) I-129 1.16 x 10 1.6 x 10 y 2.9 x 10 2.6 x 10 (3) <0.000 -8 7 6 1-131 25,200 S.1 days 6.3 1.44 x 10 5.7 1-132 44,000 2.3 hr 11.0 1.5 x 10 0.103 I 5 I-133 62,000 21.0 hr 15.5 3.1 x 10 3.00 1-134 64,000 53.0 min 16.0 3.3 x 10 0.03 I-135 69,000 6.7 hr A7.3 6.6 x 10 0.71 1 9.5 Rems i to thyroid Isotope Ci T 1/2 x 0.01 (4) Rems Xe-131m 200 12.0 d 2.0

1. 56 x 10-

<0.000 Xe-133m 1,400 2.3 d 14.0 3.3 x 10- <0.000 Xe-133 53,000 5.3 d 530.0 3.5 x 10 ' O.011 l -3 Xe-135m 15,000 16.0 min 150.0 3.4 x 10 0.029 -3 Xe-135 59,000 9.2 hr 590.0 1.9 x 10 0.064 4 -3 Xe-137 59,000 4.2 min 590.0 1.51 x 10 0.051 -3 Xe-13S 59,000 17.0 min 590.0 9.2 x 10 0.31 Kr-83m 4,300 1.86 hr 1.1 1.9 x 10 <0.000 Kr-85m 13,500 4.4 hr 135.0 1.2 x 10-0.009 Kr-85 191 10.0 y 1.9 1.72 x 10- <0.000 Kr-87 25,000 16.0 min 250.0 6.17 x 10-0.088 i -2 Kr-88 30,000 2.8 hr 300.0 1.52 x 10 0.260 Kr-89 34,000 3.2 min 340.0 1.73 x 10 ' O.336 1 1.15 Rems to whole body (1) From Radiological llealth llandbook, Dangerous Properties of Industrial Materials i Sax, N.I. 1963, and by direct calculation using reported fission vields. i (2) Rads /Ci released at 280 meters = (C1 released)(Factor *)(3.17)(X/Q). Regulatory Guide 1.109 " Calculations of Annual Doses to Man frem Routinc Releases i of Reactor Ef fluents f or the rurpose of Evaluating Coupliance with 10 Crn 50", Appendix 1, October 1977. Rev. 1, pp 1.109-21. I i (3) CRC llandbook of Radioactive Nuclides, pp 222. l (4) Calculat ed from equation D = 74fTE/m Ibid, pp 204. Report of JCRP Committec II on Permissible Dose for Internal Radiaticn (1959). I 5-23 1 i

DECDIBER, 1978 REFERENCES Meteorology and Atomic Energy 1968, USGPO, TID-24170, Figure A.4 (1968). " Report of Committec I' on Permissible Dose for Internal Radiation," Health Physics, Vol 3, June 1960. 3 Dose Conversion Factors Taken from " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, J. J. Didanno, R. E. Baker, F. D. Anderson, and R. L. Waterfield (1962). Recommendations of the International Commission on Radiological Protection, " Report of Committee II on Permissible Dose for Internal Radiation (1959)," ICRP Publication 2 (New York: Permagon Press, 1960). 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. Ibid, p 410. 9 .' Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant," USFAC, Regulatory Guide 3.34. Ibid, pp 10, 11. 11 " Hazards Evaluation in Support of Request for Amendment 6, License R-47 to Operate the Lynchburg Pool Reactor at 1000 kW," BAW-74, Supplement 8, Babcock & Wilcox Co., April 1962. 5-24

DECEMBER, 1978 6.0 EFFLUENT A'!D ENVIRONMENTAL MEASUREMENTS 6.1 PREOPERATIONAL ENVIRONMENTAL PROCPAv.S Environmental monitoring prior to construction and operation of the first facility in 1956 was not performed. 6.2 OPERATIONAL MONITORING PROCPM S_ 6.2.1 Radiological Monitorine Procram 6.2.1.1 Effluent Monitoring Airborne effluents that are potentially contaminated are exhausted through the 50 meter stack, where practicable. This stack is sampled continuously. Sample air is drawn through a fixed filter which is routinely changed and counted on a low background, gas flow proportional counter to determine gross alpha and beta activity. The sensitivity of this counting -17 -1 system is 8 x 10 Ci/ml for gross beta and 6 x 10 pCi/ml for gross alpha for the present counting period. Airborne effluents that cannot practicably exhaust through the 50 meter stack are individually sampled if there is the potential for these streams to contain 107. or greater of the applicable 10 CFR 20 limits. These samples are counted as described above. Liquid sampling is performed on each of the waste water tanks prior to discharging to the liquid waste treatment system at the Naval Nuclear Fuel Division. Tanks are stirred and a one-quart sample withdrawn. A measured amount of this sample water is evaporated to dryness on a planchet and counted in a low background, gas flow proportional counter for gross alpha and gross beta. The sensitivity of this system is 3 x 10- pCi/ml for gross alpha and 3 x 10- pCi/ml for gross beta for the present counting period. Gamma spectroscopy is used for isrtope identification if the gross technique results in unusually high activities. 6.2.1.2 Fnvironmental Monitoring The James River is sampled periodically both upstream and down-stream of the NNFD discharge point (see Fig. 6.1). Samples are evaporated to dryness on a planchet and counted on a low background, gas flow pro-portioncl counter. Samples are counted to determine gross alpha and gross 6-1

DECEMBER, 1978 beta. The sensitivity of this system is 1.5 x 10~ uCi/ml for gross alpha -9 and 3 x 10 pCi/ml for gross beta, for the present counting period. Samples of James River silt and plant life in the vacinity of the LRC are periodically taken (see Fig. 6.1). These samples are normally analyzed by an off-site commercial firm. Rain water is continuously sampled on site. Measured amounts are evaporated to dryness and counted on a low background, gas flow proportional counter for gross alpha and gross beta. The sensitivity of the system is 1.5 x 10-pC1/ml for gross alpha and 3 x 10~ pCi/ml for gross beta, for the present counting period. 6.2.2 Chemical Monitoring The liquid efflents from the LRC that potentially contain harmful chemicals are released to the liquid waste treatment system of the Naval Nuclear Fuel Division. That division analyzes effluents to chemical con-stituants and therefore this is not performed by the LRC. 6.2.3 Meteorological Monitoring Wind speed and direction monitors are mounted at the top of the 50 meter stack and at a point about midway up the stacks. The information transmitted from these monitors is recorded on a continuous basis. Outside air temperature is measured and recorded continuously at locations on and near the stack at elevations of 50 meters and 3 meters. 6-2}}