ML20133G088
| ML20133G088 | |
| Person / Time | |
|---|---|
| Site: | 07001201 |
| Issue date: | 12/31/1996 |
| From: | Elliott G FRAMATOME COGEMA FUELS (FORMERLY B&W FUEL CO.) |
| To: | Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| TAC-L30828, TAC-L30836, NUDOCS 9701150131 | |
| Download: ML20133G088 (139) | |
Text
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W 1/Wft F RAMATOM E COG EM A F U E LS December 31,1996 Robert C. Pierson, Chief Licensing Branch, NMSS Division of Fuel Cycle Safety & Safeguards, NMSS United States Nuclear Regulatory Commission Mail Stop T8D14 Washington D.C., 20555
Dear Mr. Pierson:
Reference:
Docket No. 70-1201, SNM-1168 The Demonstration Section of the above referenced license has been revised and is attached for your review. B&W Fuel Company has been changed to Framatome Cogema Fuels (Reference New Letter of Credit and Name Change: TAC No. L30828 and L30836). The Commercial Nuclear Fuel Plant has been changed to Lynchburg Manufacturing Facility. Additionally, organizational changes have been made at the facility.
Six cor
, of the Demonstration Section are included.
Because of changes throughout the Section, Part 11 is being submitted in its entirety. The changes in the Section are indicated by a side-bar and are detailed in Attachment 1.
If you have any questions regarding the revised Demonstration Section, please feel free to call me at (804) 832-5202.
Sincerely, FRAMATOME COGEMA FUELS Lynchburg Man facturin acility
.o gg 09 Gay! F. Elliott 9701150131 961231 PDR ADOCK 07001201 B
PDR Framstome Cogoma Fuels j
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T Page 1 of 5 Chan. Pana h
Change l
9 9-1 9.1 B&W Fuel Company changed to Framatome Cogema Fuels. Changed partnership names.
9.1.1 Changed B&W Fuel Company to Framatome Cogema Fuels. Changed Commercial Nuclear Fual Plant to
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Lynchburg Manufacturing Facility.
9.1.2 Changed C.W. Pryor to J.R. Bohart. Changed B&W Fuel Inc. to Virginia Fuels Inc. Changed Virginia Fuels Inc. to Cogema Fuels Inc. Changed R.H. Idhe to R.B. Hoffman. Changed B&W Fuel Company to Framatome Cogema Fuels.
9-2 9.2 Changed B&W Fuel Company to Framatome Cogema
- Fuels, 9.3 Updated throughput ranges per most recent data.
i 9.4 Changed B&W Fuel Company to Frematome Cogema i
Fuels. Changed CNFP to LMF.
9.4.1 Changed B&W Fuel Company, Commercial Nuclear Fuel Plant to Framatome Cogema Fuels Lynchburg Manufacturing Facility.
9-2,3 9.4.2 Changed CNFP to LMF.
9-3,4 9.4.4 Changed CNFP to LMF 9-4 9.4.5 Changed CNFP to LMF.
9-5 9.5 Changed CNFP to LMF.
9.6 Changed CNFP to LMF.
10 10-1 10.1 Changed CNFP to LMF. Changed BWFC to FCF.
10.1.1 Corrected references identified in accordance wtyh Figure 10.1.
1 10-1,2 10.1.2 Corrected references identified in accordance with Figure 10.1.
f Page 2 of 5 Chan. Page h
Change 10 10-2 10.1.3 Corrected references identified in accordance with Figure 10.1.
10.2.2 Chenged CNFP to LMF.
l 10-3 10.2.3 Changed CNFP to LMF.
l 10-8 10.4.1 Changed CNFP to LMF.
10-9,10 10.5.1 Changed CNFP to LMF.
10-10 10.5.2 Changed CNFP to LMF.
10-11 10.5.5 Changed CNFP to LMF.
10.5.6 Changed B&W to FCF.
10-12 10.6.1 Changed CNFP to LMF.
10-13 10.6.3 Changed CNFP to LMF.
11 11-1 11.1 Changed CNFP to LMF.
11.2.1 Changed CNFP to LMF.
11.2.2 Changed CNFP to LMF.
11-3 11.2.6 Changed CNFP to LMF.
11.3 Changed CNFP to LMF. Changed Manager of Safety and Licensing to G.F. Elliott. Changed Manager of Facilities and Services (Acting) to C.W. Carr.
11-4 11.3 Changed Commercial Nuclear Fuel Plant to Lynchburg Manufacturing Facility. Changed B&W Fuel Company to Framatome Cogema Fuels.
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11-7 11.3 Changed B&W Fuel Company Commercial Nuclear Fuel Plant to Framatome Cogema Fuels, Lynchburg Manufacturing Facility.
Attachment _1 Page 3 of 5 Chan. Page P_ata.
Change 11 11-9 11.3 Changed B&W Fuel Company Commercial Nuclear
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Fuel Plant to Framatome Cogema Fuels, Lynchburg l
Manufacturing Facility.
l 11-11,12 11.3 Added G.F. Elliott's resume in lieu of K.S. Knapp's.
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11-13 11.3 Changed B&W Fuel Company, Commercial Nuclear
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Fuel Plant to Framatome Cogema Fuels, Lynchburg j
Manufacturing Facility.
j 11-14 11.3 Changed B&W Fuel Company, Commercial Nuclear 1
Fuel Plant to Framatome Cogema Fuels, Lynchburg Manufacturing Facility.
l 11-16-18 11.3 Added C.W. Carr's resume in lieu of W.T. Foot's.
j 11-19 11.3 Changed B&W Fuel Company to Framatome Cogema l
Fuels.
11-20 11.3 Changed B&W Fuel Company to Framatome Cogema Fuels.
i l
l l
11-28 11.5.3.2 Changed CNFP to LMF.
i 11-31 Fig.11.1 Changed CNFP to FCF, LMF. Changed W.T. Foot to C.W. Carr. Changed K.S. Knapp to G.F. Elliott.
i 11-32 Fig.11.2 Omitted BWNT to President. Changed B&W Fuel l
Company to FCF.
f 12 12-1 12.1 Changed CNFP to LMF.
l 12.2 Changed CNFP to LMF.
12.2.1 Changed CNFP to LMF.
)
12-2 12.4.1 Changed CNFP to LMF.
i 13 13-1 13.1 Changed CNFP to LMF.
14 14-1 14.1 Changed CNFP to LMF.
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i Page 4 of 5 Chan. P_aga Earl Change 14 14-1 14.3 Changed CNFP to LMF.
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1 14-3 14.5 Changed CNFP to LMF.
14-4 14.6 Changed CNFP to LMF.
14-5 14.6.1 Changed B&W to FCF.
15 15-1 15.1 Changed CNFP to LMF.
15-2 15.2.2 Changed CNFP to LMF.
15-13 15.2.4 Changed CNFP to LMF.
15-13,14 15.2.4.2 Changed CNFP to LMF.
15-21 15.2.6 Changed CNFP to LMF.
j 15-28 15.2.7 Changed CNFP to LMF.
15-30 15.2.8 Changed CNFP to LMF.
16 16-1 16.1 Changed CNFP to LMF.
16-1,2 16.1.1 Changed CNFP to LMF.
16-2 16.1.2 Changed CNFP to LMF.
16-4 16.1.2.2 Changed CNFP to LMF.
16-5 16.1.2.3.1 Changed CNFP to LMF.
16.1.2.3.2 Changed CNFP to LMF.
16-5,6 16.1.2.3.3 Changed CNFP to LMF.
16-6 16.1.2.3.4 Changed CNFP to LMF.
16-7-9 16.1.2.3.5 Changed CNFP to LMF.
16-11 16.1.2.3.6 Changed CNFP to LMF.
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Page 5 of 5 t
l Chan. Page h
Change i
16 16-12 16.1.2.4 Changed CNFP to LMF.
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FRAMATOME COGEMA FUELS - LYNCH URG MANUFACTURING FACKITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH SAFETY DEMONSTRA TION - TABLE OF CONTENTS CHAPTER DESCRIPTION PAGE REVISION 9
General Information 9-1 through 9-10 3
10 Facility Description 10-1 through 10-15 4
11 Organization and Personnel 11-1 through 11-32 9
1 12 Radiation Protection 12-1 through 12-10 3
13 Environmental Safety 13-1 through 13-4 2
- Radiological 14 Nuclear Criticality Safety 14-1 through 14-10 4
i l
l 15 Process Description & Safety 15-1 through 15-37 5
Analysis 16 Accident Analysis 16-1 through 16-13 3
Page: 10-1 December 30,1996 Revision: 4
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' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTII - CHAPTER 9 - GENERAL INFORMATION 1
9.1 Coroorste information Framatome Cogema Fuels (FCF) has constructed facilities at the i
Lynchburg, Virginia site for the purpose of fabricating fuel assemblies for commercial utility nuclear reactors. Framatome Cogema Fuels is a l
partnership of Cogema Fuels Inc. and Virginia Fuels Inc. that is organized and exists under the laws of the State of Delaware.
9.1.1 Addresses of Princloal Offices l
Framatome Cogema Fuels P. O. Box 10935 Lynchburg, Virginia 24506-0935 Framatome Cogema Fuels Lynchburg Manufacturing Facility P. O. Box 11646 Lynchburg, Virginia 24506-1646 9.1.2 Coroorate Officer Information Names &
Office Citizenship Addresses J.R. Bohart President USA P.O. Box 10935 Virginia Fuels Inc.
Lynchburg, VA.
24506-0935 M. 3. McMurphy President USA l
P. O. Box 10935 Cogema Fuels Inc.
Lynchburg, VA.
24506-0935 R.B. Hoffman President USA P. O. Box 10935 Framatome Cogema Fuels Lynchburg, VA.
24506-0935 Page: 9-1 December 30,1996 Revision: 3 n
' FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY l
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USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTH - CHAPTER 9 - GENERAL INFORMATION j
1 9.2 Financial Onnlification Framatome Cogema Fuels will, from time to time at a frequency acceptable l
to NRC, provide a summary of its financial condition sufficient to demonstrate conclusively its ability to pay the cost of decommissioning its nuclear facility. Such summary will include.the method by which, and extent to which, provision is being made for this financial obligation.
1 9.3 Summarv of Ooaratino Objective and Process The utilizes uranium oxida in pellet form enriched to a maximum of 5.1%,
j UO, pellets are received and loaded into cladding material. The loaded l
rods are then configured into fuel assemblies for use in commercial power reactors. Plant throughput typically ranges between 150 to 400 l
MTU/ year. This is based on historical data. The previous license renewal was performed in 1990. Major plant process changes made since then can be described as follows:
1.
expansion of facility for non-licensed material i
2.
addition of downloading operations 9.4 Site Descriotion The following are brief summaries of certain physical attributes of the LMF l
site. The majority of the information expressed was extracted from the environmental reports prepared for Framatome Cogema Fuels (formerly B&W Fuel Company, Commercial Nuclear Fuel Plant) in 1974 and 1976.
9.4.1 Geography
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Framatome Cogema Fuels Lynchburg Manufacturing Facility is located on a 76 acre site in Campbell County, Virginia, approximately 4 miles from the Lynchburg City limits. The site is adjacent to the Babcock and Wilcox NNPD and Nuclear Technology Center plant sites. The physical layout of the LMF site is as shown l
in Figure 9.1. The entire 525 acre site is illustrated in Figure 9.2.
9.4.2 Democraohv Because of the terrain, most of the population within a 5 mile radius of the LMF resides over 3 miles from the site. There are no Page: 9-2 December 30,1996 Revision: 3
i FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 9 - GENERAL INFORMATION significant clusters of population within a 2 mile radius of the LMF.
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l The closest inhabitants occupy residsinces which are located about i
one-half mile to the ENE.
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Approximately two-thirds of the population within 5 miles of the plant reside between the 3 and 5 mile radil in the WSW to WNW directions. This includes the eastern portions of the City of l
Lynchburg and the community of Madison Heights.
Within a 3 mile radius of the plant, there are only a few public facilities or business activities that attract large numbers. The neighboring Archer Croek Plant of the Lynchburg Foundry and the NNFD and Lynchburg Technology Center of Babcock & Wilcox are the only other major industries in the immediate vicinity.
Approximately 3000 workers are employed at these facilities.
9.4.3 Meteorologv Since Lynchburg is situated in the valley of the James River and on the eastern edge of the Blue Ridge Mountains, extreme weather in the area is rare.
Severe weather at the Mt. Athos site is generally limited to occasional thunderstorms, with a very low probability of tornadoes.
According to methods for estimating tornado occurrence, the probability of a tornado actually striking the site in any given year 4
is 3.0 x 10, with a recurrence interval of 3333 years. Since the site is not a coastal location, the effects of hurricanes would be limited to increased rainfall and possible flooding.
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9.4.4 it/ rology d
The LMF site lies on a river bend bounded on three sides by the James River and on the southeastern side by Mt. Athos. Hence, the only waters that could be affected by plant operation or that could influence plant operation are the James River and the ground waters of the site and its immediate environs.
Facilities at the Mt. Athos site utilize several wells to obtain groundwater. These wells are situated in the northeastern portion of the property along the James River. Water levels at these wells Page: 9-3 December 30,1996 Revision: 3
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 9 - GENERAL INFORMATION are approximately 465 to 442 feet Mean Sea Level (MSL) and Indicate the gene' rally northerly downward-sloping trend toward the river.
1 The site groundwater supply is stored in three 150,000 gallon tanks located on a hilltop on the property immediat,ely north of the one million gallon service water tank. The LMF utilizes groundwater from these tanks at an average rate of 2500 gallons per day. Hence, the LMF accounts for a very small percentage of the total consumption of groundwater at the Mt. Athos site.
9.4.5 Seismologv The central Appalachian region of Virginia is characterized by a moderate amount of low-level earthquake activity which appears as somewhat isolated " clusters" of seismic energy release; there is a central Virginia cluster, a western Virginia-West Virginia cluster and a norther Virginia-Maryland-West Virginia cluster. The LMF is located in a western part of the central Virginia cluster region which is classified as Zone 2 on the Seismic Risk Map of the United States. On the Modified Mercalli (MM) scale this zone corresponds to an intensity of Vil, which implies building damages to the extent of fallen chimneys and cracked walls. Zone 2 has an acceleration range of 0.065 to 0.14 gravity.
9.4.6 Geologv The James River Basin of Virginia includes portions of four physiographic provinces, each of which is characterized by distinct land forms and physical features. These provinces, located west to east, are Valley and Ridge, Blue Ridge, Piedmont and Coastal Plain.
Western or inner Piedmont where t.he Mt. Athos site property lies is in upland characterized by scattered hills, some of mountainous dimensions, lying eastward from the foot of the Blue Ridge.
j The Mt. Athos site is located on a river bend and generally exhibits a rolling surface of gentle slopes it is bounded on three sides by the meandering James River and on the southeastern side by Mt.
Athos. The dominant topographic feature of the site is a hill located approximately at the center of the property, the crest of which rises to 693 feet MSL. The ground is inclined toward the Page: 9-4 December 30,1996 Revision: 3
' FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 9 - GENERAL INFORMA TION river from the hilltop to the river bank, which is at approximately 470 feet MSL. The highest point in the vicinity of the site is the top of Mt. Athos, where the elevation is 890 feet MSL.
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9.5 Location of Buildinos on Site The physical layout of the LMF site is illustrated in Figure 9.1. The relationship of the LMF to other facilities at the Mt. Athos site is shown in Figure 9.2.
9.6 Maps The following maps are included to clearly define the boundary of the site and the relationship to neighboring areas as well as within the State of Virginia. These are as follows:
Points of Interest in the Vicinity of the LMF l
Figure 9.3 Physical Features Within Five Miles of the LMF l
Figure 9.4 Figure 9.5 The Relatioriship of the LMF to Major Virginia Population Centers 9.7 License History The following is a chronology of SNM-1168 from initialissue to the present:
Dale Descriotion 12/69 SNM-1168 first issued 4/76 First renewal of SNM-1168 6/83 Second renewal of SNM-1168 5/88 Expiration date of SNM-1168 9/90 Renewal of SNM-1168 granted for a ten year period Page: 9-5 December 30,1996 Revision: 3
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FRAMA TOME COGEMA FUELS - L YNCHBURD MANUFACTURING FACILITY i
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PAfsTll - CHAPTER 9 - GENERAL INFORMATION
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Page: 9-6 December 30,1996 Revision: 3
' FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 9 - GENERAL INFORMATION 1
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Page: 9-7 December 30,1996 Revision: 3
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTl/ RING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 9 - GENERAL INFORMATION FIGURE 9.3 I
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USNRC LICENSE SNM-1968, DOCKET 70-1201 1
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Page: 9-9 Decernber 30,1996 Revision: 3
1 FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 9 - GENERAL INFORMA TION FIGURE 9.5 8 ',.
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Page: 9-10 December 30,1996 Revision: 3
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 10 - FACILITY DESCRIPTION 10.1 Plant Lavout and Goerations Figure 10.1 is a layout of the LMF illustrating the various production areas on sit's. The plant's primary function is the manufacture of nuglear fuel assemblies for use in commercial power reactors. These operations may i
be subdivided into three production phases: unclad SNM Handling, Fuel Rod Processing and Inspection, and Fuel Bundle Assembly. There are occasions when the fuel has to be downloaded. For downloading operations, the phases are reversed. The numbers in parenthesis are taken from Figure 10.1.
The LMF also supports the Field Operations Department for the l
refurbishment of contaminated equipment. The operations associated with FCF, mainly fuel related, are performed in the south bay of the main plant, l
known as the Service Equipment Refurbishment Facility (SERF-1). A small hot machine shop is adjacent to the main facility and is referred to as SERF-2. Refurbishment operations for the liquid volume reduction and chemical cleaning processes are performed in the SERF-3. SERF-4 is the largest of the SERF facilities and is located on the back of the property behind the main manufacturing plant. Primary operations to be conducted at SERF-4 include the decontamination, maintenance, testing and storage of contaminatea equipment, tooling and components associated with various field service activities.
10.1.1 Unclad SNM Handling Unclad SNM receiving, storage, and rod loading are located at the south end of the plant, as shown in Figure 10.1. The area includes pellet receiving (#1), the pellet vault (#5), and the pellet loading room (#7). Other than the laboratory (#11), this is the only part of the process in which unciad special nuclear material (SNM) is handled. The entire pellet vault / rod loading area is separated from the remainder of the plant by means of concrete block and metal walls. A slight negative pressure is maintained in this area with respect to the rest of the plant to prevent contamination spread.
10.1.2 Fuel Rod Processinn.and insoection Loading fuel rods are processed and stored in the central portion of the plant (#8). Processing includes end cap welding, product quality inspection, cleanin0 (#15), helium leak testing, and Page: 10-1 December 30,1996 Revision: 4
' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1968, DOCKET 70-1201
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PARTll - CHAPTER 10 - FACILITY DESCRIPTION l
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accumulation of rods into groups of the number required for a fuel assembly. Rods are then stored until needed for assembly production. Individual unciad fuel pellets are processed in the j
laboratory (#11) which is located in this portion bf the plant.
l j
i 10.1.3 Fuel Bundle Assembiv l
i Fuel rods are assembled into their final configuration (#12),
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checked for quality, and shipped to the customer from the north i
end of the plant (#13).
l Ancillary production activities conducted within the plant consist of non-nuclear component fabrication which may be characterized as light machining and fabrication. Examples of this type of activity include grid and end cap production, incore detectors, and dimensional adjustment on vendor-supplied components.
}
10.2 Utilities and Suonort Svstems 10.2.1 Electric Power 4
i Electric power to the Mt. Athos site is provided by Appalachlan Power Company. This power is supplied via a nearby electrical i
substation and is stepped down to 480V 3-phase,3-wire service.
A further step down to 240V,120V, and 277V is made for lighting and general convenience power.
Backup battery power is provided for the criticality alarm, fire alarm and public address in addition to emergency lighting. The nature of
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our operations is such that a loss of utilities simply results in a l
totally safe halt in operbtions.
t 10.2.2 Comoressed Air I
Compressed air is utilized primarily far routine industrial purposes.
We do not use any protective masks or clothing that require compressed air to maintain their effectiveness. We have a main compressor located at the north end of the LMF that provides the compressed air for plant use. A desiccant is used to dry all plant i
compressed air.
j Page: 10-2 December 30,1996 Revision: 4 d
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' FRAMATOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 10 - FACIOTY DESCRIPTION 10.2.3 Water The Mt. Athos site utilized several wells on site to obtain ground water. The ground water supply is' stored in two 150,000 gallon tanks. Additionally, 2 - one million gallon storage tanks are maintained for service water. Typically, the LMF uses l
approximately 2500 gallons per day.
l A loss of water supply would not lead to any degradation of our j
safety systems or contribute to.an accident that could release l
uranium to the plant or the environment.
l 10.3 Ventilation Svstems 10.3.1 General Airborne contamination will be maintained as far below 10 CFR 20 Appendix B limits as is practicable. Containment and isolation of areas where unciad SNM is processed in significant quantities is assured by enforcing pressure differential criteria so that such areas are negative with respect to the remainder of the plant. Air circulation within controlled areas is maintained by the use of a combination of fresh " makeup" and filtered air. The relative percentages of fresh and recycled air will be determined by air handling and tempering requirements, for example, air conditioning.
Recycled air is routed through a pre-filter and is HEPA filtered before return to the operating area. Determination on the necessary number of air changes per given period of time will be based on design criteria and health physics operational experience.
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10.3.2 Overall Svstem Design Figure 10.2 is a schematic which illustrates the configuration of the controlled area ventilation system including the relative location of sampling points, pre-filters, HEPA filters and the effluent release point. Certain design criteria have been established and maintained for this system as follows:
l 1.
Individual HEPA filter units are installed at the rate of 1 l
filter /1000 CFM of air flow, or more if allowed by filter l
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' FRAMATOME COGEMA FUELS - LYNCHBURG ALANUFACTURING FACillTY USNRC LICENSE SNM-1968, DOCKET 70-1201 PARTM - CHAPTER 10 - FACILITY DESCRIPTION 1
specifications (typical HEPA filter specifications are shown in Table 10.1).
l 1
2.
Effluents exhausted to the environment shall be.HEPA l
filtered.
3.
Self-closing louvers are installed at outside air intake points to the south bay.
4.
HEPA filter banks are contained in metal units specifically designed to allow:
Access to space between filter banks to allow in-place monitoring for defects.
Removal and replacement of filters from outside the housing structures with the use of " bag-out" techniques for contamination control.
Measurement of pressure drop.
5.
Pre-filters will be used to limit duct contamination and to provide protection to the HEPA filters if necessary.
Selection of single or dual prefiltration is based on filter loading potential.
6.
Duck construction will be metal with sealed mechanical joints where practicable. Connections to containment units may be fabricated of flexible or semi-flexible material with joints bolted or fastened by an equally effective technique.
7.
j Recirculated air is first passed through pre-filters and is then HEPA filtered before reentering the area.
8.
Provision for DOP testing and sampling is incorporated in the ventilation system design.
Page: 10-4 December 30,1996 Revision: 4
' FRAMATOME COGEMA FUELS - LYNCHBURG ALANUFACTURING FACILITY USNRC LICENSE SNM-1968, DOCKET 70-1201 l
PARTll - CHAPTER 10 - FACillTY DESCRIPTION 3
}
TABLE 10.1 f
l TYPICAL FILTER SPECIFICATIONS 5
Manu. &
Filter' Frarne Rated Effic.
Capacity Max.
Model Media Material CFM Temp.
HEPA Cambridge Glass Steel or 99.97 1100 2507 i
ET-1000-1 24" x Wood 0.3 u l
24" x 11.5" l
Pre-filters Glass Galvanized 60 %
1000 2507 l-Cambridge Steel or ASHRAE (AEROPAC) 3CP-Wood Std. 5268 l
60-24246 23 3/8" x 23 3/8" x 5 7/8" i
J Flanders Glass Pressed 95% NBS 2000 j
(Econocel) 2424F, Board
" Dust Spot" 23 3/8" x 23 3/8" x Test i
11 %"
i i
l i
\\
i 1
l 1
1 i
Page: 10-5 December 30,1996 Revision: 4
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURINO FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTM - CHAPTER 10 - FACillTY DESCRIPTION i
10.3.3 Overall Swatam Maintenance and Control l
Pressure drop across pre-filter systems will be monitored and the filter replaced when the pressure differential reaches 4_ inches of water. Calculations by a consultant indicate that, for the purposes I
of nuclear safety, pressure drop measurements will provide j
effective control over the quantity of SNM that may accumulate in a filter. Calculatioas show a " worst case" P of 2 inches of water for a 2.72 kg UO, t?ter load, assuming a P.of.5 inches H,0 for a new filter, a UO density of 5 g/cc, and for conservatism, a linear 2
l function between P and load.
i l
Further scaling indicates that for a P of 4" H,0, the total i
accumulation of UO, would be 5.44 kg. This is equivalent to 200 l
grams of assU or approximately 25% of the 850 g safe mass limit.
The pre-filters are therefore inherently safe and may be handled for maintenance and cleaning, etc., within the handling limits imposed for nuclear safety purposes.
l Typical pre-filter and HEPA filter housings are designed by the manufacturer to accommodate safe bag-out techniques for i
contaminated filters. In order to accomplish the above, air flow d
through the filter is shutoff and the housing cover plate removed.
l A plastic bag may then be attached to the filter housing port, the i
filter housing release handles loosened, and the filter withdrawn l
Into the bag which can be sealed and removed from the housing.
l The bag-out technique, applied as needed, effectively precludes significant airborne and surface contamination spread.
l Flow sensing elements are installed within ductwork at appropriate locations to provide an audible and visual alarm if flow is interrupted. If flow loss is other than a momentary disruption, operations will be terminated if necessary until proper ventilation is restored.
Pressure sensing devices are routinely monitored to assure that the corstrolled area remains negative with respect to the remainder of the plant.
j Air recirculated back into the controlled area is sampled on a continuous basis to verify filter effectiveness and will not be Page: 10-6 December 30,1996 Revision: 4 4
r
l
- FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURIND FACKITY USNRC LICENSE SNM-1968, DOCKET 70-1201 PARTM - CHAPTER 10 - FACKITY DESCRIPTION recirculated if the levels are above 25% MPC of 10 CFR 20 Appendix B.
Procedures require notification of Safety & Licensing pr.ior to
servicing or maint'enance on those portions of the fans or HEPA systems that may be contaminated.
10.3.4 Individual Onarations containrnant 1
i The degree and type of containment required for individual l
operations is determined based on the dust generation potential of j
the activity being conducted. Three generalized categories of J
containment have been utilized in developing the airborne exposure control systems as follows:
1.
Glove boxes, locally fabricated or commercially available, i
for use in areas of high potential contamination spread.
2.
Containment hoods are designed to accommodate specific equipment or operations and incorporate specific measures to minimize the potential for contamination spread including:
Minimization of normally open penetrations to those a.
necessary for routine operation. (NOTE: In some cases no "normally open" access will be required.)
- b..
Penetrations that may be required for infrequent non-routine activities will be sealed by "normally g
closed" covers at all times when not in use.
3.
Standard fume or " chemical type" (open face) hoods for use in operations with minimal potential for dust generation and to accommodate analytical activities if needed. As appropriate, measures, such as blocking of doors, will be utilized to minimize open face areas.
10.3.5 Individunt Containrnent - Deslan Criteria The following criteria have been established for the design and construction of individual containment devices:
Page: 10-7 December 30,1996 Revision: 4 l
l i
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1968, DOCKET 70-1201 i
PART11 - CHAPTER 10 - FACILITY DESCRIPTION
}
i 1
j a.
Design face velocity shall be 100 LFM with all ports open.
j The air handling system incorporates the capability to j
adjust face velocity as necessary in order to satisfy exposure control needs.
i b.
in-line pre-filter units will be installed in the downstream air flow path and as near to the contamination source as is practicable based on design and operating criteria.
]
c.
A negative pressure of approximately 0.25 inches of water relative to room ambient, will be maintained in glove box units.
l d.
Materials of construction will be nonflammable or flame
}
retardant. Structural numbers and those portions of the containment device not requiring visibility will be fabricated i
j of metal. Where visibility is needed, plexigla'ss, satisfying l
at a minimum, the "self-extinguishing" criteria of ASTM D635 (" Flammability of Self-Supporting PlasJcs") or 1
equivalent will be used. Tempered glass, such as is l
normally used in standard fume hoods, is also acceptable.
l Flammable materials such as polyethylene are not authorized.
10.3.6 Individual Containment - Maintenance and Control l
The adequacy of installed systems is verified by an air sampling l
program during the startup phase of the operations and necessary
)
modifications or design changes made to assure that operator i
exposure is as low as reasonably achievable. In addition to the I
above, local capture devices (" elephant trunks") could be used to l
provide added support as necessary.
l 10.4 Radioactive Waste Handling-10.4.1 L! auld Wastes 1
Potentially contaminated liquids generated at the LMF are
{
controlled by means of dedicated evaporation systems for all i
radiological controlled areas. The liquid effluent is collected and 4
allowed to evaporate (with heat if necessary) into the existing i
1 3
Page: 10-8 December 30,1996 Revision: 4 i
i
_____._____._.__._._____m b,'
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACiUTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PART11 - CHAPTER 10 - FACiUTY DESCRIPTION i
airborne effluent control system where it is HEPA filtered prior to release. The HEPA system and HEPA filtered prior to release. The HEPA system and 10 CFR 20 airborne effluent release limits used are as described in 8.1.1. Vessels used to collect / evaporate the liquid effluent shall be inspected monthly for sludge accumulation.
Any dried sludge or other solids collected from the i
holding / evaporation vessels will be disposed of as LSA waste.
l 1
10.4.2 Solid Wastes I
Uncontrolled disposal of solid wastes or equipment is authorized when contamination levels do not exceed the levels defined in i
section 1.7.4 and under the concept of ALARA.
Establishment of the above contamination limits to permit disposal i
in accordance with routine industrial practice does not present a hazard to the general public. The limits are generally accepted l
within the nuclear industry, as not presenting any significant i
radiological or nuclear safety hazards.
i l
Routine monitoring programs are conducted by Safety & Licensing to assure that material, contaminated in excess of specification limits, is not released for uncontrolled disposal and to detect and alleviate increasing contamination trends.
i i
Non-contaminated solid wastes are disposed of through a contract hauler. Contaminated solid wastes consist primarily of low specific activity material and are disposed of by a licensed contractor by j
land burial on an NRC or state licensed site. LSA wastes are i
packaged in appropriate containers as required by 10 CFR and 49 l
CFR. SNM content for each package is estimated using gamma scan or by gross alpha count.
l 10.5 Fire Protection 1
j 10.5.1 General 1
f All LMF buildings are of steel and/or masonry construction and the j
roofs of all main buildings are Class I construction. Class I construction requires that the vapor barrier be non-combustible.
i Plant operations are typical of metal working type facilities; j
l Page: 10-9 December 30,1996 Revision: 4
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' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 10 - FACILITY DESCRFTION i
1 l
therefore, very few Class A type combustibles are present.
i Accumulations of combustible materials within the LMF shall be limited to the greatest extent practicable, consistent with operational requirements. Supervision is responsible for assuring that areas under their cognizance are maintained in accord with j
good housekeeping and fire prevention practice. Plant operating i
procedures coupled with frequent inspections deter sloppy i
housekeeping which would allow the accumulation of combustibles
{
in the work area. Safety & Licensing personnel are responsible for j
inspecting and maintaining fire protection equipment.
10.5.2 Im& mentation of the Fire Protection Procram i
i LMF has several documents that implement our fire protection l
program. Internal Safety & Licensing procedures are used to ensure employees are properly' trained on fire safety, to outline training for the LMF fire brigade, to.equire routine inspection of I
l fire protection and emergency equipment to meet NFPA standards, j
to control welding and other " hot" work, to provide emergency l
response in the event of a fire, and to mandate independent audits i
of our fire protection program. Our pre-fire plan is an extensive document that provides a hazard analysis for all areas of the plant.
Each area includes an assessment of the occupancy, access and egress, lighting and communication, fixed fire systems and manual j
i suppression systems, ventilation system, and guidelines for the i
i attack and special precautions. Hazards associated with each area l
are also identified which include hazardous substances with material safety data sheets, NFA rating and quantities, physical i
hazards, electrical hazards, and compressed gases.
i
\\
10.5.3 Fire Extinguishing Svstems l
j Fire extinguishing systems compatible with area nuclear safety l
requirements shall be installed or provided in accord with insurance l
and federal regulations. The above systems s5all include portable j
extinguisher of a type (pressurized water, CO and dry chemical) 2
{
and size based on the potential hazard. Agents such as "Metl-X" j
are available in areas where metal fires may occur.
1 I
l Page: 10-10 December 30,1996 Revision: 4
' FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACLU1Y
)'
. USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 10 - FACILITY DESCRIPTION l
Automatic sprinkler systems are installed in accord with standard industrial practice when compatible with nuclear safety and operational requirements.
9 Sprinkler systems and other water-type extinguishing systems are not installed in moderation controlled areas. Only approved fire fighting techniques are to be used in moderation control areas.
These approved methods shall be prominently posted in the affected areas.
10.5.4 EJammable Llaulds Flammable liquids shall be stored in containers that are approved t'y FM or UL or are otherwise acceptable to the insurance carrier.
10.5.5 Fire Brigade l
As part of its fire protection program, the LMF organization shall include a " fire brigade" staffed by qualified personnel, familiar with j
basic fire fighting techniques, the equipment available for their Immediate use onsite, and the nuclear safety and health physics considerations that are involved. Fire Brigade members shall be retained at least annually.
10.5.6 Offsite Sunnort Arrangements have been made for assistance from local area fire i
departments. They have been informed of FCF's operations, materials, and characteristics of potential fires and have toured the facil} ties.
l 10.5.7 S-1 Storage Facility The storage of flammable, combustible, or reactive liquids in the S-1 building will be strictly prohibited. No burning, welding or other ignition will be permitted. If welding or such other hot working operations are mandated, a pre-operational evaluation shall be performed under the direction of the Safety Review Board chairman and shall be monitored for compliance until the operation has been officially terminated. Smoking will be prohibited at all times.
Page: 10-11 December 30,1996 Revision: 4
}
' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACRJRING FACillTY l
USNRC LICENSE SNM-1968, DOCKET M-1201 PARTM - CHAPTER 10 - FACillTYDESCRIPRON l
i 10.5.8 SERF-3 & 4 l
The buildings denoted as SERF-3 and SERF-4 incorporates a sprinkler system installed throughout the facility to NFPA 13 code.
1 l
The alarm for the sprinkler system is audible locally and at the l
guard house which is manned continuously.
f Portable fire extinguishers are available throughout the building in accord with NFPA 10. Fire watches shall be conducted while performing hot working operations and controlled by the use of the local safety rules.
i All systems shall be maintained in accord with NFPA codes.
10.6 Chemical Safetv 10.6.1 General
~
l LMF differs from other fuel fabricators as it is not licensed to perform any chemical operations which involve special nuclear material such as the conversion of UF. to UO.and scrap recovery.
2 Chemical usage at LMF is for cleaning purposes. Acetone, TCE, l
hydrofluoric and nitric acid compromise the majority of the i
chemicals used at LMF. With the exception of acetone, the chemicals are used in our cleaning room which is located within the i
main plant. Acetone is used throughout the plant.
10.6.2 Chemical Umage and Onantity Acetone is used as a cleaning agent. All the containers are fire proof and the largest container has a 2 gallon capacity. Several of these containers are located in each area. Designated locations are set up to ensure proper disposition of the waste. The cleaning room has five sinks that contain acid solution for a total volume of about 70 gallons. A 120 gallon dispensing tank located in a room adjacent feeds the sinks. There is also a 400 gallon tank that collects the spent acid. Both tanks are constructed in accordance with vendor specifications for the contents it retains. They also meet the VA Department of Waste Management Regulations which mandates that they be constructed to detect leaks and requires daily inspections. A chemical storage building is located fifty feet Page: 10-12 December 30,1996 Revision: 4 i
i
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY LSNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 10 - FACILITY DESCRIPTION away from the main plant. It houses about 500 gallons of acetone and/or alcohol and about 300 gallons of acid. The acid is stored separately in an area that is diked and has an explosion proof wall.
l The LMF is licensed to store UF. cylinders. Since the.UF. is strictly for storage, there are no chemical hazards associated with the material.
t 10.6.2.1 Chemical Usage and Quantity - SERF-4 i
Stators (motors) are dipped into a varnish solution.
Toluene is introduced as required to dilute impregnated varnish to reach optimum viscosity.
Once the mixture is d: lute, the stators are placed into an oven and the coating is baked onto the stators. On occasion, toluene.is wiped onto the stators for cleaning purposes.
The average annual usage of toluene is 25 gallons per year. The average annual usage of varnish is 20 gallons per year. Both chemicals will be received in 5 gallon containers and stored in flammable storage cabinets in the SERF-4 Facility.
10.6.3 Chemical Accident l
At worst, the chemicals stored at LMF could spill. As illustrated above, this would result in a minor spill of minute quantities. Due to the low quantities of chemicals involved and the fact that the chemicals are located an adequate distance from SNM, a spill
)
would not pose a threat to cause spread of contamination.
LMF has several employees tralned in hazmat and NNFD has agreed to provide back up services. Safety & Licensing has an internal procedure to respond to chemical spills.
Page: 10-13 December 30,1996 Revision: 4
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURIN3 FACIUTY l
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 10 - FACillTYDESCRIPTION
[
l l
FIGURE 10.1 puurfsetnf h
4
-5 N
(
4 0
PELLET RECErv NG t_;
t s
U i:==:1 4
SEftF-3 S.
PELLET YAULT 40 6 METLM M M 37
\\
7 ff.LLET LGADING RM
/
8.
FUEL R00 F AB.
9.
EAST OFF8& S b
' 7
&E L
LA8.
7 d
~I
[
\\
- 12. FUEL DJOLE ASSEheLY
- 13. FUEL SWDLE STORAOE I
\\
- 14. ION EXCHANDE/ GRO GRNOINO IS. CLEANING ftOOM l
\\
- 16. FINAL FUEL 3sSPECTION l
- 87. GRO AftEA
\\
- 18. MAINTENANCE / ELECTRICAL SHOP r
l 26
\\
- 19. FaELO OPERATK)NS TRAILER l
- 20. MACHIPE SHOP 24 I-
/
38 l
j
/
- 2. W H SA Y
FICE i
29 l
- 23. ACO WASTE STORAGE iliiil1 r l1lllI11
---)'
l q
- 24. S I BUILDING l
2S. 62 BUILDtNO 35 s 3D l
[
32
- 26. 33 SUILDfMG 23 %
Ol Ig 28
- 27. FUEL ASSEheLY SHIPPING 38 20 I
- 28. TFTR 8UILDING 3
25 l10l 74 l
- 30. TRA N NO CENTER
- 29. LSA SCAN BueLDING l0g 8 li'5 6l 17 l 21 S t. RECOROS STORAGE I
5 t-d 7 l 22 %
i t2 l13 3$*
I l
H OFFICES g
34 3
- 34. EMERGENCY EOuiP. BUILDINO gL y-I 33 l
- 35. RECEIVING / STOCKROOM BUILOINO
)
u
- 38. MAIN CohPRESSOR SUILOING lc 3 l
- 37. SERF.3 C06FRESSOR Bud. DING lg W
- 38. CtfMICAL STORAGE BUfLDING
)
9 AY WASTE ACCUwlVLATION FACILITY SOUTH b
P ING POND NORTH I
PARKING LOT MUSTER 2 AREA FA\\92\\0RW\\92013 irii B-SK-BSLO60895 GENERAL LICENSING INFORMATION Page: 10-14 December 30,1996 Revision: 4
I FRAMA TOME bOGEMA FUELS - L YNCHBURG MANUFACTURIND FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 10 - FACILITY DESCRIPTION FIGURE 10.2 FLOW DIAGR"AM OF AIR HANDLING SYSTEM EFFLUENT h
SAMPLE POINT (*8000 CFM)
HEPA FILT9ATION I
u 1
i 1
-1 1
PRE-FILTER I
I SAMPLE I
POINT I
I CONTROLLED AREA ROOM AIR PICKUP I
I RECIRCULATING SYSTEM i
1 I
I I
ROOM AIR 4 HEPA I
RETURN FILTRATION i
I
^
l I
h PRE-FILTER PRE-FILTER a
A PROCESS ROOM AIR HOODS PICKUP Page: 10-15 December 30,1996 Revision: 4
~.
' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACiUTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL l
11.1 Ornanl7ational Reanonsibilities 1
Figure 11.1 illustrates the departmental and managerial organization at the LMF. The key organization responsible for maintaining the health and j
safety aspects at the LMF to include all SERF Facilities is the Health-Safety 1
Section. The Health-Safety Section is a part of the Safety and Licensing i
organization. The Health-Safety Section reports to the Manager, Safety and Licensing. The Manager, Safety and Licensing reports directly to the j
Plant Manager.
l 11.2 Kev Personnel Function l
11.2.1 Overall Program Management 1
Responsibility for planning, coordinating, administering and managing the health and safety aspects of the LMF is vested in the l
Manager, Safety and Licensing. This position is organizationally parallel to other members of the Plant Manager's staff such as the Managers of Manufacturing Engineering and Fuel Manufacturing.
11.2.2 The Health-Safety Section Health-Safety personnel are responsible for the general surveillance of all plant safety related functions. Specifically, these functions are described as follows:
Maintaining appropriate control of hazardous material, shipments, and receipts.
Supervising and coordinating the hazardous waste disposal program.
Assisting in personnel and equipment decontamination.
Distribution and processing of personnel monitoring equipment.
Maintaining individual exposure records.
Orienting and training LMF personnelin radiological and nuclear safety.
Page: 11-1 December 30,1996 Revision: 9
FRAMA TOME C00EMA FUELS - LYNCHBUR3 MANUFACTURING FACM.lTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 8
Furnishing consulting services and advice on radiation 4
protection to plant supervision and management.
Generating or acquiring, mainthining, and appropriately distributing all records and reports required by applicable regulations or procedures.
Leak testing on sealed radioactive sources.
Developing and disseminating procodures related to radiation safety and reviewing procedures prepared by other operating sections for regulatory compliance and the adequacy of safety considerations.
The key positions within the Health-Safety Section are the Health Physicist, the Regulatory Compilance Officer and the Health-Safety Monitors.
11.2.3 Health Physicist The Health Physicist is responsible to provide management with assurance of the effectiveness of the entire health and safety program from a radiological, nuclear, industrial, and chemical safety aspect. This position is responsible for evaluating the potential for exceeding authorized control limits and to recommend appropriate restrictions or corrective measures.
The Health Physicist is also responsible for supervising the implementation of the Health-Safety Program to assure that the requirements as defined by license and procedures are carried out.
The Health Physicist reports directly to the Manager, Safety &
Licensing.
11.2.4 Regulatorv Comoliance Officer The Regulatory Compliance Officer is responsible for implementing the occupational and industrial safety programs to include chemical and fire safety. The Regulator Compliance Officer reports directly to the Manager, Safety & Licensing.
Page: 11-2 December 30,1996 Revision: 9
~
- FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 11.2.5 Health-Safetv Monitors The Health-Safety Monitors are responsible for conducting routine monitoring, sample collection and analytical tests to determine radiation and contamination levels. The Health-Safety Monitors l
report directly to the Health Physicist.
11.2.6 SERF 3 & 4 Organization i
The SERF 3 & 4 facilities have a Site Manager to oversee the operations of both facilities. His staff includes 2 supervisors and maintenance technicians. The SERF 3 & 4 Manager reports to the Manager, Chemical & Environmental Services who reports to the Manager, Integrated Nuclear Services who reports to the BWNT
)
President, CEO. This is illustrated in Figure 11.2 A health physicist and 2 health safety technicians from the LMF l
Safety & Licensing organization are also permanently assigned to the SERF 3 & 4 facilities.
11.3 Resumes Since it is also the responsibility of the entire plant management to assure safe operations and regulatory compliance, resumes from other managerial organizations within the LMF are also included. These are as fol'ows:
l Name litle C.W. Carr Vice-President, Manufacturing & Services (Plant Manager)
J.T. Ford Manager, Fuel Manufacturing (Production Manager)
T.S. Wilkerson Manager, Manufacturing Engineering (Production Manager)
G.F. Elliott Manager, Safety & Licensing G.B. Lindsey Health Physicist D.L. Gordon Sr. Health Physicist C.W. Carr Manager, Facilities and Services (Acting)
L.A. Hassler Senior Principal Engineer P.L. Holman Senior Principal Engineer F.M. Alcorn Manager, Nuclear Criticality Safety Engineering J.M. Harwell Nuclear Criticality Specialist Engineer Page: 11-3 December 30,1996 Revision: 9
FRAMA TOME COGEMA FUELS - L YNCNBUR3 MANUFACTURING FA CIUTY
\\
USNRC LICENSE SNM-1168. DOCKET 70-1201 PARTH - CHAPTER 11 - ORGANIZA110N AND PERSONNEL i
NAME:
Charles W. Carr TITLE:
Vice-President, Manufacturing and Services 4
(Plant Manager, Lynchburg Manufacturing Facility)
CITIZEN OF UNITED STA TES EDUCATION: Virginia Polytechnic Institute and State University, Blacksburg, Va.
BS in Mechanical Engineering - 1966 Registered Professional Engineer in Virginia Reg. No. 8442 EXPERIENCE: 1991 - Present Framatome Cogema Fuels (formerly B&W Fuel Company), Lynchburg Manufacturing Facility (formerly Commercial Nuclear Fuel Plant),
Lynchburg, Va. - Plant Manager. Responsible for all plant operations including safety, licensing, safeguards, environment, manufacturing, and quality.
1986-1991 Babcock & Wilcox Co., Nuclear Power Division.
Lynchburg, Va. - Manager, Component Engineering & Field Services. Responsible for technical adequacy of modifications and new construction'of components in nuclear steam systems serviced by B&W. Also responsible for field services including new equipment installation, plant modifications, refueling, and consultation.
l Section duties include estimating, proposal, customer interfaces on technical issues, and 4
contract execution within authorized budget and schedule. Supervisory responsibilities for staff of five managers and a group of 80 to 120 engineers and technicians.
1983 -1986 Babcock & Wilcox Co., Nuclear Power Division, Lynchburg, Va. - Manager, Engineered Products &
Parts. Provided hardware for several major repair / service projects in B&W and W nuclear steam systems. Improved revenue in the EP&PS Business Unit and improved the quality of Page: 11-4 December 30,1996 Revision: 9
i FRAMATOME COGEMA FUELS - LYNCHBUR3 MANUFACit/RWG FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZATION AND PERSONNEL hardware and the service to B&W nuclear customers.
1981 -1983 Babcock & Wilcox Co.', Nuclear Power Division',
Lynchburg, Va. - Project Manager, Nuclear Engineering Services. Managed several service contracts ranging from $20K to $4.2M including design and procurement of new reactor coolant pump seals for five B&W and two CE designed nuclear steam systems. Responsibilities included cost control, estimating, customer interface and vendor contract negotiations. Negotiated several agreements for technology transfer, both to and from EdF in France.
1976-1981 Babcock & Wilcox Co., Naval Nuclear Fuel Division, Lynchburg, Va. - Unit Manager, Core Structurals Procurement. Managed a group of engineers responsible for specification and procurement of core structurals (non-fuel) for naval reactors. Provided technical evaluation and j
disposition of manufacturing discrepancies, engineering changes and manufacturing procedures. Responsible for vendor technical and schedule performance.
1972-1976 Babcock & Wilcox Co., Naval Nuclear Fuel Division, Lynchburg, Va. - Senior Engineer, Advanced Core Engineering. Developed j
manufacturing processes for an advanced NR fuel design. Designed tooling and purchased (specified) welding equipment. Completed pre-production hardware and final pre-production report.
1968-1972 Babcock & Wilcox Co., Naval Nuclear Fuel Division, Lynchburg, Va. - Design Engineer in Core Assembly Section. Wrote procedures for assembly, testing, and shipment of nuclear fuel for the U.S. Navy. Obtained customer (U.S.
Page: 11-5 December 30,1996 Revision: 9
_.. _. _ _ _ _ _ _. =. _ _ _. _ _ _ _ _ _ _.
FRAMA TOME COGEMA RIELS - L YNCHBUR3 MANUFACTURIN3 FACMJTY USNRC LICENSE SNM-1168. DOCKET 70-1201 PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL 1
1 1
Government) approval of all procedures and engineering changes.
i l
1967-1968 Combustion Engineering, Chattanooga,.Tenn.,
Erection Department - Erection Site Field Engineer l
for large controlled circulation and combined i
circulation steam generators. Worked with~ erector l
and craft labor to resolve erection problems, make i
plans, and monitor schedules.
i i
l 1
l 4
i i
i i
4 I
l i
i i
Page: 11-6 Decernber 30,1996 Revision: 9
i FRAMA TOME COGEAGA FUELS - L YNCNBUR3 MANUFACTURING FACKITY USNRC LICENSE SNM-1968, DOCKET 70-1201 i
PARTH - CHAPTER 11 - ORGANIZATION AND PERSONNEL i
NAME:
J. T. Ford i
T/TLE:
Manager, Fuel Manufacturing i
l CITIZEN OF UNITED STATES I
~
i i
l l
EDUCAT/ON: Davidson County Community College j
A.A. - 1969 Central Virginia Community College, Lynchburg, Va.
Nuclear Technology - 1975 l
M/L/TARY:
1970 - 1972 U. S. Army l
EXPER/ENCE: 1985 - Present Framatome Cogema Fuels, Lynchburg i
Manufacturing Facility, Lynchburg, Va. - Manager, Fuel Manufacturing. Responsible for fuel rod, i
contro! component, grid, fuel bundle, and incore detector fabrication.
1982-1985 B&W Fuel Company, Commercial Nuclear Fuel Plant, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Regulatory Control.
4 Responsible for nuclear materials control, licensing, health-safety, physical security, nuclear l
safety, receiving and stores.
i 1981 - 1982 Babcock & Wilcox, Commercial Nuclear Fuel Plant, l
Lynchburg, Va. - License Administrator.
)
Responsible for obtaining and administering all licenses and permits including those issued by NRC, EPA, State, and Local agencies.
1979 - 1981 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Foreman, Bundle Assembly.
Responsible for assembly operations, and the shipment of finished products.
i 1978 - 1979 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Foreman, Grid Manufacturing.
Responsible for grid manufacturing.
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1975 - 1978 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Foreman, Pelletizing Operations.
Responsible for pellet operations.
~
1972 - 1975 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Data Evaluation Technician.
Reviewed and certified fuel assemblies and control 2
components.
1 1970 -1972 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Manufacturing Technician.
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h d
s j6 i
d
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACMJTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL NAME:
T. Scott Wilkerson i
T/TLE:
Manager, Manufacturing Engineering 3
i l
CITIZEN OF UNITED STATES i
EDUCAT/ON: University of Virginia, Charlottesville, Va.
M.E. In Applied Mechanics - 1986
)
Virginia Polytechnic Institute and State University, Blacksburg, Va.
]
B.S. In Mechanical Engineering - 1978 l
l 5
EXPER/ENCE: 1992 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, j
Manufacturing Engineering. Unit manager with j
responsibility for manufacturing procedures and processes used for production of nuclear fuel l
components. Unit provides technical responsibility for equipment design, fabrication methods, j
process qualifications and evaluation of deviated components. Function includes liaison with j
vendors, fuel design engineering and customers.
I 1990 -1992 B&W Fuel Company, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Group Supervisor in charge of the Present Production Engineering Group of i
the Fuel Mechanical Engineering Unit. As such, responsible for the interface between the design of fuel assemblies and control components and the fabrication facility. All drawings, specifications, l
j and associated documentation for component j
fabrication are released, maintained, and administered by this group.
l 1980-1990 B&W Fuel Company, Lynchburg, Va. - Principal j
Engineer, entered as an Associate Engineer (Engineer ll). Work consists primarily of the mechanical (structural) design testing, analysis, and associated documentation of commercial i
nuclear fuel assembly structural components and related equipment. This work typically involves the use of open shop computer programming as 1
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j well as large scale structural analysis computer l
codes such as ANSYS. Project Administrative tasks such as preparing proposals, schedules, presentations, cost estimates, and technical reports are routinely done as necessary. Often this work is in close cooperation with the j
manufacturing plant; therefore, familiarity with standard manufacturing practices is also maintained.
1978-1980 Lynchburg Foundry, Lynchburg, Va. - Associate Engineer. In this capacity responsible for the-budget, schedule, design, installation, and startup i
of capital projects and for solving a variety of l
mechanical plant engineering problems. Most of the work was in direct support of the
{
manufacturing and quality control process of the l
plant. Lynchburg Foundry makes gray and ductile Iron castings primarily for the automotive and j
heavy equipment industries.
i l
PROFESSIONAL l
AFFILIATION:
ASME l
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i 1
i 1
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' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACKITY l'
USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PART11 - CHAPTER 11 - ORGANMA TION AND PERSONNEL 1
NAME:
Gayle F. Elliott 1
i l
TITLE Manager, Safety & Licensing l
CITIZEN OF UNITED STATES
~
EDUCATION: Liberty University, Lynchburg, Va.
j M.B.A. - 1990 j
B.S. In Mathematics - 1988.
l l
EXPERIENCE 1996 - Present Framatome Cogema Fuels, Lynchburg l
Manufacturing Facility, Lynchburg, Va. - Manager, Safety & Licensing. Responsible for coordinating
{j the technical aspects of radiation control for the fuel manufacturing plant and for the field operations refurbishment facility. Involved in the jecommissioning, training, emergency l
preparedness, and transportation prograrns.
j Responsible for budgeting for the Safety &
j Licensing organization. Liaison with regulator j
agencies.
i 1993-1996 Framatome Cogema Fuels, Lynchburg j
Manufacturing Facility, Lynchburg, Va. - Technical Specialist lil, Quality Technology. Cognizant l
engineer for inspection processes, procedures and equipment for burnable poison, control rod, axial l
power assemblies, receipt inspection components, l
fuel and burnable poison pellets. Support Safety &
Licensing with uranium accountability and j
l licensing. Facilitator for Field Services Quality j
Improvement Teams and Quality Steering Team.
1992-1993 B&W Fuel Company, Commercial Nuclear Fuel l
Plant, Lynchburg, Va. - Group Leader, Manufacturing Engineering. Supervise activities in i
the Manufacturing Engineering Machining and l
Special Processes group which includes directing i
the work for individuals within the group and assuring that milestones and schedules are met for i
such. Ensure complete, accurate and high quality i
l l
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j documentation of all contracts. Provide technical review of documents to ensure all aspects of l
design criteria are met. Administer presentations j
for customers and employees of manufacturing i
processes. Coordinate R&D projects and prepare
{
budget report for manhours and material costs.
t i
1989 -1992 B&W Fuel Company, Commercial Nuclear Fuel i
Plant, Lynchburg, Va. - Technical Specialist II, Manufacturing Engineering: Cognizant process 4
engineer for incore Detector Assemblies and j
related components. Responsible for weld and braze qualifications, upgrade of fabrication
{
procedures routing documents, establishment of i
research and development projects to support j
growth and progress, and troubles' hooting.
Provide engineering support of all BWNS related i
contracts and ensure implementation of the applicable sections of the ASME Code. Originate and administer procurement of enriched and
]
natural pellets.
1 l
1988 -1989 Educational Leave for of Graduate Studies.
i i
1983-1988 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Engineering Aide, Manufacturing Engineering. Responsible for preparing documents for ordering contract materials, i.e. Purchased l
Materials Lists, Bills of Materials and Engineering *
)
Requirements. Initiate Route Cards and Contract i
Submittal Requirements. Assist in Echo 330 ultrasonic fuel inspection. Perform administrative duties during refueling outages, i'
1981 -1982 Babcock & Wilcox, Nuclear Power Division,
}]
Lynchburg, Va. - AA Tech Co-op, Systems Engineering. Responsible for assisting organization in various engineering assignments and producing
)
engineering schedules and work authorization j
reports.
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' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACKETY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL NAME:
Gerald B. Lindsey
~
TITLE:
Health Physicist CITIZEN OF UNITED STATES l
EDUCA TION: Virginia Polytechnic Institute and State University, Blacksburg, Va.
i B.S. In Biology - 1975 j
CVCC Emergency Medical Technician - 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> j
(Coordinated through Blue Ridge Emergency Medical Service) l EXPERIENCE: 1994 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Health Physicist and Training Instructor. Additional duties include training on radiation protection.
a j
1986-1994 Framatome Cogema Fuels, Lynchburg j
Manufacturing Facility, Lynchburg, Va. - Health-i Safety Foreman. Duties include standard review i
and implementation, safety training program, plant i
safety audits, procedure writing, accident investigations.
i 1985 - 1986 Babcock & Wilcox Company, Commercial Nuclear l
Fuel Plant, Lynchburg, Va - Senior Health-Safety i
Monitor.
j 1983 - 1985 Babcock & Wilcox Company, Lynchburg Research l
Center, Lynchburg, Va. - H.P. Surveyor for the
)
Building C Decommissioning Project.
4 1976-1983 Babcock & Wilcox Company, Commercial Nucisar l
Fuel Plant, Lynchburg, Va - Quality Assurance Lab l
Technician and Health-Safety Monitor.
1969-1976 Lynchburg General Hospital Emergency Aoom, Lynchburg, Va. Duties include vital signs, emergency aid, patient care.
MEMBERSHIPS / CERTIFICATIONS:
i i
National Registry for Radiation Protection Technologists - 12/93 l
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USNRC LICENSE SNM-1168. DOCKET 70-1201 l
PARTH - CHAPTER 11 - ORGANIZA TION AND PERSONNEL I
i j
NAME:
Darryl L. Oordon i
j TITLE:
Sr. Health Physicist
~
CITIZEN OF UNITED STATES i
i EDUCAT/ON: University of Lowell, Lowell, MA j
B.S. In Radiological Sciences - 1988 EXPERIENCE: 1991 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Sr. Health j
Physicist. Responsible for Coordination, Management and Technical Development of the l
Radiation Protection Program for fuel manufacturing and field operations refurbishment divisions. This responsibility includes radiological
~
training, emergency response coordination, compliance with NRC, EPA, State, and Local regulations, procedure development, Bioassay, and j
ALARA Program, 1988 - 1991 Portsmouth Navhl Shipyard, Portsmouth, NH -
Radiological Control Technician. Responsible for i
Radiation and Contemination control associated i
with the overhaul, maintenance, and refueling of i
Naval nuclear propulsion systems. Established j
guidelines for various industrial tradesmen to reduce exposure, minimize radwaste, and prevent l
spread of contamination to uncontrolled areas, l
personnel, tools and equipment. Ascisted in l
Processing and disposal of waste, man-rem j
estimates, procedure development, environmental impact monitoring, decommissioning, and 4
radicactive material shipments.
1986-1987 Pennsylvania Power and Light Company, Allentown, Pa. - Health Physicist. As Cooperative Associate, performed various activities to assist the corporate Health Physics organization in support of an operational two-unit BWR (Susquehanna SES) including resolution of l
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACM11Y USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTN - CHAPTER 11 - ORGANIZATION AND PERSONNEL l
i 1
l operational HP problems, ALARA review, LLRW i
storage facility safety analysis, primary system i
gamma spectroscopy, and man-rem estimates.
Also coordinated the activities of the company's l
by-products materials license including source management and inventory, regulatory j
compliance, and instrument calibration services.
l 1985 Applied Healt'h Physics, Inc., Holyoke, Ma. -
l Health Physics Technician. Performed cleanup i
operations of a large manufacturing facility l
following a fire involving Polonium-210.
l Responsible for air, water, soil, and surface l
contamination surveys and removal of j
j contaminated plant equipment. Packaged and shipped Radioactive-LSA waste and interacted with regulatory officials (NRC/ EPA) to monitor the i
total environmental impact of the fire.
1 MEMBERSHIPS / CERTIFICATIONS:
i, Certified Radiological Control Technician -
(NAVSEA 389-0288 Art.108)
I i
j 4
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' FRWAATOn6E COGEMA FUELS - LYNCHBURG RCANUFACTURIN3 FACillTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTN - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 1
i NAME:
Charles W. Carr (Acting) i l
T/TLE:
Manager, Facilities and Capital CITIZEN OF UNITED STATES EDUCAT/ON: Virginia Polytechnic Institute and State University, Blacksburg, Va.
BS in Mechanical Engineering - 1966 Registered Professional Engineer it. Virginia Reg.No.8442 EXPER/ENCE: 11/96 - Present Framatome Cogema Fuels, Lynchburg (Acting)
Manufacturing Facility, Lynchburg, Va. - Manager, Facilities and Capital. Responsible for operation and maintenance of plant, grounds, and equipment. Plans additions to buildings and plans for efficient layout of equipment.
1991 - Present Framatome Cogema Fuels (formerly B&W Fuel Company), Lynchburg Manufacturing Facility (formerly Commercial Nuclear Fuel Plant),
Lynchburg, Va. - Plant Manager. Responsible for all plant operations including safety, licensing, safeguards, environment, manufacturing, and quality.
1986-1991 Babcock & Wilcox Co., Nuclear Power Division, Lynchburg, Ya. - Manager, Component Engineering & Field Services. Responsible for technical adequacy of modifications and new construction of components in nuclear steam systems serviced by B&W. Also responsible for field services including new equipment installation, plant modifications, refueling, and consultation.
Section duties include estimating, proposal, customer interfaces on technical issues, and contract execution within authorized budget and schedule. Supervisory responsibilities for staff of five managers and a group of 80 to 120 engineers and technicians.
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. USNRC LICENSE SNM-1168, DOCKET 70-1201
)
PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL l
l 1983 -1986 Babcock & Wilcox Co., Nuclear Power Division, Lynchburg, Va. - Manager, Engineered Products &
i Parts. Provided hardware for several major repair / service projects in B&W and W nuclear steam, systems. Improved revenue in the EP&PS Business Unit and improved the quality of hardware and the service to B&W nuclear customers.
1981 -1983 Babcock & Wilcox Co., Nuclear Power Division, Lynchburg, Va. - Project Manager, Nuclear Engineering Services. Managed several service contracts ranging from $2OK to $4.2M including j
design and procurement of new reactor coolant 4
pump seals for five B&W and two CE designed j
nuclear steam systems. Responsibilities included j
cost control, estimating, customer interface and j
vendor contract negotiations. Negotiated several j
agreements for technology transfer, both to and l
from EdF in France.
1976-1981 Babcock & Wilcox Co., Naval Nuclear Fuel i
Division, Lynchburg, Va. - Unit Manager, Core i
Structurals Procurement. Managed a group of i
engineers responsible for specification and procurement of core structurals (non-fuel) for i
i naval reactors. Provided technical evaluation and i
disposition of manufacturing discrepancies, engineering changes and manufacturing i
i procedures. Responsible for vendor technical and i
schedule performance.
1 i
1972-1976 Babcock & Wilcox Co., Naval Nuclear Fuel Division, Lynchburg, Va. - Senior Engineer, Advanced Core Engineering. Developed manufacturing processes for an advanced NR fuel design. Designed tooling and purchased l
(specified) welding equipment. Completed pre-j production hardware and final pre-production j
repod.
i i
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 1968-1972 Babcock & Wilcox Co., Naval Nuclear Fuel Division, Lynchburg, Va. - Design Engineer in Core Assembly Section. Wrote procedures for assembly, testing, and shipment of nucliar fuel for the U.S. Navy. Obtained customer (U.S.
Government) approval of all procedures and engineering changes.
i 1967-1968 Combustion Engineering, Chattanooga, Tenn.,
Erection Department - Erection Site Field Engineer for large controlled circulation and combined circutation steam generators. Worked with erector and craft labor to resolve erection problems, make plans, and monitor schedules.
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l FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURIN1) FACRJTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTH - CHAPTER 11 - ORGANIZA TION AND PERSONNEL i
NAME:
L. A. Hassler i
TITLE:
Senior Principal Engineer i
CITIZEN OF THE UNITED STATES l
EDUCAT/ON: University of Va., Charlottesville, Va.
)
Ph.D. In Nuclear Engineering - 1973 l
St. Louis University, St. Louis, Mo.
B.S. In Physics - 1965 j
)
EXPERIENCE: Framatome Cogema Fuels, Lynchburg, Va. - Senior Principal
)
Engineer in the Fuel Engineering Section with 20 years experience i
in the areas of radiation transport, shielding, and criticality safety.
j in the area of criticality safety he has been involved in the j
following tasks:
Attending ORNL KENOlV training course and KENOVa training course.
Analysis of the disrupted core at TMI-il and analytical design of the l
TMI-il defueling canisters using KENOlV program.
i On loan for three months to the Navel Nuclear Fuel Division (NNFD) criticality safety group. Supported criticality safety analysis for l
storage at NNFD.
l Provided criticality design analysis for the BR-100 spent fuel l
shipping container using KENOlV program.
I
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Provided criticality analysis for the NNFD SX22 shipping container j
SARP license submittal.
Provided criticality analysic for the new fuel storage racks for i
Wisconsin Public Services Corporation and for the spent fuel l
storage racks for Toledo Edison Corporation.
Provided QA review for the licensing criticality calculations for two i
types of BWFC new fuel shipping containers.
I
)
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USNRC LICENSE SNM-1168,' DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL i
NAME:
P. L. Holrnan TITLE:
Senior Principal Engineer l
CITIZEN OF THE UNITED STATES
~
i EDUCATION: University of Va., Charlottesville, Va.
M.S. in Nuclear L :gineering - 1978 j
B.S. In Nuclear Engineering - 1975
~
EXPERIENCE: Framatome Cogema Fuels, Lynchburg, Va. - Senior Principal l
Engineer with 17 years experience in reactor physics, including the following areas:
j Core Follow analysis, Standard Models Development, Feel Cycle Division for Oconee Units I, il, lil, ANO-1 Unit 1, and Crystal River, Licensing, Reload Licensing Analysis Task Engineer, Maneuvering analysis, and Criticality calculations using Monte Carlo l
methodology.
Recent criticality work involved licensing the ANF/BWFC Model
)
51032-2 new fuel shipping container and also included separate
{
criticality analyses for new and spent fuel storage racks for NUSCo l
and the Wisconsin Public Services Corporation. Preceding that i
work was the criticality analysis for the Toledo Edison new and j
spent fuel storage racks which involved an enrichment limit increase utilizing burnup credit. Additional criticality work involved j
the analysis for the GPU TMl-2 defueling canisters and the DOE j
100 ton rail-barge shipping cask.
t i
TECHNICAL PAPERS AND PUBLICATIONS:
)
"TMI-2 Defueling Canisters Reactivity Analysis", ANS Transactions, j
1986.
l
)
"Three-Dimensional Analyses of a Highly Heterogeneous PWR l
Using the NOODLE Code", ANS Transactions 1988.
PROFESSIONAL AFFillATIONS:
1 j
American Nuclear Society, Virginia Chapter ANS I
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PARTM - CHAPTER 11 - ORGANIZATION AND PERSONNEL i
)
NAME:
Francis M. Alcorn i
TITLE:
Manager, Nuclear Criticality Safety Engineering CITIZEN OF UNITED STATES
~
j EDOCATION: Lynchburg College, Lynchburg, Va.
M.B.A. - 1974 University of Va., Charlottesville, Va.
M.E. In Nuclear Engineering North Carolina State University, NC B.S. In Nuclear Engineering - 1957 EXPERIENCE: 1957 - 1960 Babcock & Wilcox, Atomic Energy Division, Lynchburg, VA. Functioned as a Nuclear Engineer doing both core neutron physics and shielding calculations.
1960-1961 General Nuclear Engineering Corporation, Staff Physicist. Engaged in core neutron physics design and analysis of the Bolling Nuclear Superheat Reactor. Wrote physics articles for Power Reactor Technolotiv. which was published by GNEC for the AEC at that time.
1961 -1964 Babcock & Wilcox, Nuclear Power Generation Division, Lynchburg, VA. Concerned with core neutron physics analysis and design of the Consolidated Edison Reactor, the Liquid Metal Fuel
)
Reactor, the Babcock & Wilcox Test Reactor, the Advanced Test Reactor, the Heavy Water-Organic Cooled Reactor Concept, and Babcock & Wilcox Pressurized Water Reactor Concepts. Developed methods for and performed calculations for criticality, fuel depletion, nuclear safety coefficients, power profiles, nuclear fuel costs, and critical experirnent analysis. Worked in the areas of kinetic safety analysis.
1964-1969 Babcock & Wilcox, Utility Power Generation Division (formerly Nuclear Power Generation Page: 11-21 December 30,1996 Revision: 9
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL Division), Lynchburg, VA. Physicist in the PWR Development Section and was responsible for determining the most economical method for l
utilizing plutonium as a recycle fuel in B&W's pressurized water reactor concepts. In addition, l
was Nuclear Criticality Safety Advisor to the Company's Naval Nuclear Fuel Division.
1969 -1971 Babcock & Wilcox Company, Lynchburg Research l
Ceater, Lynchburg, VA - Criticality Specialist, Nuclear Safety Engineer. Transferred to the LRC as Nuclear Criticality Safety Specialist for Babcock l
& Wilcox's Naval Nuclear Fuel Plant, Commercial Nuclear Fuel Piant, and the LRC.
4 1971 to April 1987 j
Babcock & Wilcox, Lynchburg Research Center, 3
Lynchburg, VA - Supervisor, Nuclear Criticality l
Safety Group. This group was the Company's central organization which provided guidance, j
developed and validated the analytical methods needed for criticality evaluations, did criticality
]
calculations, performed nuclear safety audits, and gave assistance to the various divisions of the Company and the Company's customers in 2
matters related to nuclear criticality safety. In 4
i addition to responsibility as supervisor of this group, was the Nuclear Safety Officer for the l
Lynchburg Research Center (LRC).
I i
April 1987 - Present Manager, Nuclear Criticality Safety Engineering.
Responsibfe for managing the Nuclear Criticality
]
Safety Engineering Unit which develops and i
validates the analytical methods needed for criticality safety evaluations, performs criticality safety calculations needed within Babcock &
Wilcox, conducts nuclear safety audits, and assists the various divisions of the Company and the Company's customers in matters related to nuclear criticality safety. This unit, within the i
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PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL i
i Naval Nuclear Fuel Division, was formerly located I
within the Research and Development Division; its responsibilities and functions remain essentially unchanged. Also the Nuclear Safety O_fficer for the Naval Nuclear Fuel Division Research j
Laboratory (formerly the Lynchburg Research Center).
1 4
1 i
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PARTM - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 2
1 l
NAME:
J. Wayne Harwell TITLE:
Principal Engineer, Nuclear Criticality Safety CITIZEN OF UNITED STA TES EDUCA TIOA!
B. S. Nuclear Engineering,1963 Mississippi State University i
M. S. Nuclear Engineering,1968 j
Mississippi State University EXPER/ENCE: 1963-1964 Ingalls Shipbuilding Corporation - Engineer, i
Shielding Structure Unit Performed nuclear shield
)
design modifications and project management related duties for shielded structures on nuclear 4
i submarines during construction.
l 1964-1968 Mississippi State University - Instructor & Graduate
]
Assistant Engineering Graphics Department.
l Graduate assistant and instructor teaching j
freshmen engineering drawing classes. Attended
]
graduate school in nuclear engineering.
1968-1976 Babcock & Wilcox, Nuclear Power Division, Senior Engineer, Nuclear Development Work related to
]
self-powered neutron detectors signal-to-power conversion, core physics analytical modeling and I
benchmarking, core model analyses, core and fuel assembly design optimization and reactor vessel l
fluence analysis.
i l
1976-1976 Southern Company Services, Senior Core Analysis Engineer. Developed core physics models for the Farley PWR cores including generation of cross j
section table sets and geometries for'PDOO7 using EPRI ARMP code package.
1976-1988 Babcock & Wilcox, Fuel Management Analysis.
}
Responsible for fuel cycle design and fuel 4
management analyses for Connecticut Yankee and l
B&W design 177 fuel assembly reactor cores using Page: 11-24 December 30,1996 Revision: 9
- FRAMA TOME.COGEMA FUELS - LYNCHBURG MANUFACTURING FACMJTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL the PDOO7 computer code. Work includes cross section table set generation and fitting strategy development, advanced fuel and reactivity control concept development, new fuel management concepts and use of transport codes for analytical model development.
1988-Babcock & Wilcox Company - Priricipal Engineer
\\,
Nuclear Criticality Safety Engineering. Performs nuclear criticality safety evaluations using the l
SCALE computer code package that utilizes the Monte Carlo computer codes (KENO-4 and KENO-
- 5) and transport computer code (XSDRN).
- Responsible for methods development along with computer codes benchmarking, verification, and validations for the codes used in nuclear criticality i
calculations.
1 1
I I
L l
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1 l
11.4 Operating Procedures
}
Written procedures for the conduct of specific operations including i
i maintenance and development of work within the plant are prepared by the j
functional component responsible for that activity. Health Safetv activities are controlled by detailed operating procedures developed by Health-Safety i
to assure standardization and accuracy. All written procedures.are reviewed and approved by appropriate representatives of plant management. If SNM or other radioactive materials are involved in,an activity, appr' oval by the Manager, Safety and Licensing or his designee shall be required prior to implementation. Likewise all Health-Safety procedures are approved by the l
Manager, Safety and Licensing as well as by affected members of plant management. Health-Safety procedures are reviewed periodically and updated accordingly.
s Applicable procedures are made available in the work area and adherence to procedure is required of all personnel.
11.5 Training All personnel receive basic training in radiological, industrial, and nuclear safety upon being hired. This initial training is a cooperative effort involving Human Resources, Health-Safety, and the employee's supervisor and is designed to satisfy the requirements of 10 CFR 19.12.
Particular emphasis is placed on the nature of the materials handled, ALARA plant safety program and rules,10 CFR 19 requirements, and the emergency evacuation system. Specific areas covered in the safety training program are as follows:
i 11.5.1 initial Ernolovee Training Employees are referred to Health-Safety by the Human Resources Department for initial training in safety. The entire plant safety program is reviewed in some detail with particular emphasis being placed on specific areas depending on the employee's job assignment. A brief discussion of, and familiarization with, the i
general principles of health physics and nuclear safety is included.
The employee is informed of his rights and responsibilities under CFR 19, and OSHA.
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USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTM - CHAPTER 11.-
ORGANIZATION AND PERSONNEL l
)
Following the initial indoctrination depending on the employees work location, they shall receive additional safety training from their immediate supervisor regarding the nuclear and radiological safety requirernants of their 6pecific job assignment. Training sessions are documented and filed as part of the employee's Health-Safety record.
i 11.5.2 Emolovee Retrainina Continuing training of a general nature is provided as necessary by Hecith-Safety and supervision. This training may be formalized i
j (i.e., " classes") or informal and conducted as part of routine Health-Safety audits. Formalized retraining may be utilized to explain operational changes affecting safety, control of special problems such as increased airborne activity, or changes in license i
specifications. The responsibility for determining the necessity for j
retraining or special training rests with Health-Safety based on plant conditions or the request of ' supervision.
l Radiation workers are all retrained annually as a routine part of the i
safety training program. The retraining sessions are documented l
and kept as part of the employee's Health-Safety record.
l l
If workers have had similar radiation worker training, the initial i
training or retraining may be by-passed by successful completion of i
a written exam with a score of 75%.
j.
11.5.3 Snecialized Training 11.5.3.1 Resolratorv Protection Training and retraining in the use of respiratory protection devices is provided by Health-Safety as required. Points relating to proper use are covered as the unit is issued and fitted by Health-Safety. This approach provides continuing review of respiratory protection requirements. Should situations arise where frequent use of a respirator is necessary, frequent Health-Safety surveillance will assure continued proper application.
Page: 11-27 December 30,1996 Revision: 9
' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 11 - ORGANIZA TION AND PERSONNEL
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11.5.3.2 Ernergencv Teams Training Specialized training for special and emergency response units such as the Fire Brigade, Radiation Monitoring Team and First Aid personnel is coordinated by Health-Safety. Fire Brigade training is conducted by representatives of Health-Safety and/or local Fire Departments and covers the use of fire fighting equipment and agents available at LMF. Radiation l
Monitoring Team members receive periodic training from Health-Safety in emergency response techniques, instrument use and maintenance, health physics and nuclear safety fundamentals, respiratory protection and contamination control. Annual evacuation drills are generally utilized as a training period for the emergency teams. First aid training is given by a qualified instructor and is the standard Red Cross program or equivalent.
11.6 Changes in Procedures. Facilities. and Eautoment 11.6.1 Procedural Changes Procedural changes are initiated.by the functional component responsible for that activity. Such procedural changes are
{
reviewed and approved by plant management prior to Implementation, if the activity involves SNM or other radioactive materials, the Manager, Safety and Licensing must approve the procedural change prior to implementation.
j 11.6.2 Facilities and Eautoment Changes Changes or modifications to facilities and equipment that have a potential impact on nuclear, radiological, industrial, or chemical safety must be reviewed and approved by the Safety Review Board and/or the Safety Review Board Chairman or qualified designee prior to initiation. The Safety Review Board is described in detail in Chapter 2.0 of Part 1.
Page: 11-28 December 30,1996 Revision: 9
' FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM 1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL 11.6.2.1 initiating Changes The responsibility for initiating changes as described in 11.6.2 is usually given to the immediate operational supervisor or manager. The requested change is documented and submitted to the Safety Review Board Chairman for initial review.
11.6.2.2 Analvsis of Changes l
The Safety Review Board Chairman determines what I
safety evaluations are needed. If the proposed modification changes the basis on which the nuclear criticality safety was originally assessed, a technical evaluation by the nuclear criticality safety group will be initiated. The organizational structure and minimum qualifications of the nuclear criticality safety group is as described in Chapter 4.0.
Radiation safety evaluations will be performed for new or revised operations to assure personnel protection is maintained. Chemical and industrial safety aspects of proposed modifications will also be evaluated for acceptability. These evaluations are documented and retained as described in 11.6.2.6.
11.6.2.3 Management Review As a minimum, the Safety Review Board Chairman or his j
qualified designee shall review all safety analyses performed the plant modifications prior to implementation. The Safety Review Board Chairman will determine if Safety Review Board approval is required.
11.6.2.4 Anoroval and Verification of Changes Approval and release of plant modifications for routine use is dependent upon satisfactory completion of a pre-operational evaluation. This evaluation is a final verification that the proposed change has been installed consistent with the analyses performed under 11.6.2.2.
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FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 11 - ORGANIZATION AND PERSONNEL 4
This evaluation will consider nucleer, radiological,.
Industrial, and chemical safety as well as license compliance. This evaluation is performed by Heal'th-Safety personnel and is approved by the Safety Review Board Chairman prior to implementation. Compliance of plant modifications is assured by our existing Health-Safety controls and audit programs with regard to i
contamination control, personnel exposures, nuclear safety, chemical and industrial hazards.
I i
11.6.2.5 Records i
All analyses, evaluations, pre-operational evaluations and other pertinent documentation relating to plant i
modifications will be maintained on file for at least six months after termination of the operation.
l i
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' FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZATION AND PERSONNEL l
FIGURE 11.1 J
1 C. W. Carr Plant Manager FCF, LMF
~'l t
T. 5. Wilkerson A. Jenkins Manager Manager Manufacturing Engineering SERF 3&4 C.W. Carr (Acting)
J. J. Bird i
Manager Manager Facilities & Capital CNFP Accounting J. T. Ford G.F. Elliott Manager Manager j
Fuel Manufacturing Safety & Licensing 3
-Plant Health Physicist
-$ERF 3&4 Health Physicist
-Training Instructor
-Regulatory Compliance
-4 Plant HP Technicians
-2 SERF 3&4 HP Technicians t
A. A. Pugh Manager Field Services Page: 11-31 December 30,1996 Revision: 9 t
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 -
ORGANIZATION AND PERSONNEL j
i i
f]GURE 11.2
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J President. CCO
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FCF Industrial Technology laternatle, mal Liason Legal Treasury & Flaance Corporate Semices Corporate Performance Integrated Nuclear Serv.
Cheafstry & Environmental l
Facilities, $ERF 3 & 4 Environmental & Waste ser Cheelstry Services Stena Generator Cleaning NuResin Services QA CIP j
Sales Force Financial Analysis 1
Procurement Outage Services NPC & Valve Services Engineering Services Plan Component Business Operation Page: 11-32 December 30,1996 Revision: 9
i FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
l PARTM - CHAPTER 12 - RADIATION PROTECTION j
i l
12.1 General i
i Operations involving potential exposure to radioactive materials will be performed in a manner that assures the radiation safety of employees and j
the general public. This policy.is implemented by maintaining a staff of l
qualified personnel and appropriate equipment, procedures, and records.
Operations will be conducted in accordance with applicable Federal, State, j
and Local requirements. Exposures to radioactive materials or other i
hazards will be maintal'ned as low as reasonably achievable (ALARA).
j Health-Safety has the authority to stop hazardous or potentially hazardous j
operations until correction or resolution by the plant management is j
obtained. Program effectiveness will not be reduced as a result of changes instituted by plantTnanagement. Typical sources of information and l
I guidance for the LMF radiation protection program are: (1) US Nuclear i
Regulatory Commission Regulatory Guides; (2) International Commission on Radiation Protection publications; (3) Handbook of industrial Loss Prevention, by Factory Mutual Engineering Corporation; and (4)
Radiological Hamith Handbook, by US Department of Health, Education, and Welfare, Public Health Service.
12.2 ALARA It is the policy of the LMF to maintain occupational exposures to radiation
]
and radioactive contamination in effluents as low as reasonably achievable.
The responsibility for implementation of the ALARA policy is designated to the Health-Safety section. Key points'that illustrate the commitment to the ALARA concept are the following:
12.2.1 The LMF uses a Safety Review Board to evaluate changes or l
modifications to facilities are equipment that have a potential impact on nuclear, radiological, industrial, or chemical safety. This assures that the changes in operations are evaluated for safety and modified as necessary before changes are implemented.
12.2.2 Surveillance, monitoring, and audit activities are executed by Health-Safety. This serves as a check of existing operations to assure compliance is maintained and safety systems.are no degraded.
12.2.3 Each employee is responsible to follow all Health-Safety procedures and other safety requirements in his work area. Employees are Page: 12-1 December 30,1996 Revision: 3 j
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5 FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURIN3 FACIUTY USNRC UCENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 12 - RADIA TION PROTECTION i
instructed to promptly report any potential safety hazards or j
incidents to the area supervisor or Health-Safety to facilitate l
prompt correction of the condition.
I 12.2.4 Each supervisor is responsible to assure that subordinaies' follow all Health-Safety requirements. Area supervisors are to follow work in j
conjunction with Health-Safety to keep all exposures as low as i
possible and keep Health-Safety informed whenever major changes l
In operational procedures, equipment, or personnel are anticipated.
12.2.5 Exposure to airborne radioactivity is maintained ALARA by engineered or other controls. Design and operational criteria of 4 j
j DAC-Hrs per shift are applied to all routine operations.
l 12.2.6 Activities that increase the probability for ingestion or Inhalation of i
hazardous or radioactive materials are prohibited.
i i
12.3 Procedures l
All licensed activities related to radiation protection shall be conducted in j
accordance with approved written procedure. Approval, scope, format, i
and distribution requirements are described in Section 2.6.
12.4 Postings 12.4.1 A continued exemption is requested from the labeling and posting i
j requirements of 10CFR20.1902(e) because of the nature of the operations at the LMF. To meet the intent of the regulatory 4
j sections cited, designated entrances to the facility are posted with a sign that contains the radiation symbol (10CFR20.1901(a)) and j
the following warning:
i j
CAUTION:
RADIOACTIVE MATERIALS. ANY AREA OR j
CONTAINER WITHIN THIS PLANT MAY CONTAIN.
RADIOACTIVE MATERIALS.
This exemption is based on practicality and experience and has been applied effectively at the LMF for the past twenty-five (25) l years.
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i FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACMJTY i'
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 12 - RADIATION PROTECTION i
12.4.2 Local safety rules approved by Health-Safety providing personnel 4
and supervision with specific directions essential to ensuring j
radiation, criticality, and industrial safety shall be posted in areas j
where appropriate.
i j
12.4.3 Other radiation safety postings or warnings as required by l
10CFR20 shall be placed in areas as required. These postings and l
warnings are used to identify such areas as " Radiologically l
Controlled Areas" (RCA), " Contaminated Areas" (CA), " Radiation j
- Areas" (RA), or "High Radiation Areas" (HRA). Transition points to 3
areas of higher radiological controls are clearly identified in the plant and discussed in training provided to all employees granted i
unescorted access to the facility.
j 12.5 Radiation Control and External Exnosure Monitorina l
Radiation areas shall be posted and controlled according to 10CFR2O l
requirements. Personnel radiation exposures shall be monitored by Health-i Safety using appropriate devices. based on the type of radiation and j
sensitivity requirements, including thermoluminescient dosimeters (TLD) or j
self-reading pocket ion chambers (SRD). Dosimetry issue to plant personnel and visitors wi'i, be determined based on the monitoring l
requirements of 10CFR20. Additional tersonnel may be monitored by l
Health-Safety for the purposes of determining exposure to various plant l
employee groups, even when these groups may not require monitoring to determine compliance with exposure limits. Accidental neutron radiation j
exposures will be monitored by indium foil issued to personnel as an integral part of the standard identification badge. Other dosimetry will be issued as necessary for unusual circumstances such as source manipul 6 tion or work involving highly transient exposure rates. Extremity j
exposure monitoring is accomplished using TLD badges as required by l
10CFR20. TLDs are processed by a vendor at monthly or quarterly intervals. Immediate processing is available for rapid evaluation of l
exposures. Personnel monitoring reports shall be prepared as required by -
l applicable regulations. Personnel exposures are reviewed periodically to l
ensure exposure levels are within regulatory limits.
l 2
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FAClllTY USNRC LICENSE SNM-1168 DOCKET 70-1201 PARTM - CHAPTER 12 - RADIA TION PROTECTION
- i 12.6 Contamination Control Prooram i
~
12.6.1 Controllad Arane j
Radiologically Controlled Areas (RCA) shall be established by i
Health-Safety as an area used to control work involving surface l
contamination above uncontrolled area limits. RCAs may contain a j
" clean area" which is potentially contaminated area and one or more Contaminated Areas (CA), which are areas known to be j
contaminated. Personnel must frisk prior to leaving a RCA or the immediate area adjacent to the RCA boundary. Upon leaving a CA, personnel must frisk prior to working in another area of the RCA.
Personnel working in a RCA shall be property trained prior to work or be escorted by a qualified worker.
I f
Contaminated Areas shall be designed to include a step-off pad or l
an intermediate change room. The purpose of the step-off paa is j
to establish a designated area where personnel enter and exit the i
Contaminated Area. Prior to stepping on,the step-off pad, all j
protective equipment and anti-contamination clothing shall be j
removed. In RCAs which do not contain a clean area (the entire RCA is a Contaminated Area),.an intermediate change room may be l
provided in lieu of the step-off pad. Guidelines for entry into j
intermediate change rooms shall be established for each area.
12.6.2 Radioloalcal Work Permits Radiological Work Permits (RWP) shall be used to define the protective clothing and equipment required to perform work l
involving surface contamination. RWPs shall be used to control all work in a RCA and work involving surface contamination above l
clean area limits outside of a permanent RCA that is not addressed by an approved operating procedure. Supervisors are required to request a RWP be written for the work to be performed. RWPs are l
written based on work area surveys and/or previously completed work and are reviewed for industrial safety prior to being approved by a Health Physicist. Work in progress is periodically reviewed to l
verify that prescribed radiological controls are adequate for the j
work being performed.
i I
1 i
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' FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACIUTY l'
USNRC LICENSE SNM-1168, DOCKET 70-1201
}
PARTll - CHAPTER 12 - RADIATION PROTECTION 1
l
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l 12.6.3 Protective Clothing 1
All persons entering Contaminated Areas (CA) will be responsible for wearing the protective clothing requirements listed on the RWP.
j Typically, protective clothing consists of a lab coat or coveralls, gloves, and shoe covers. Additional clothing may be necessary depending on the type of work being performed.
4 12.6.4 Frinkina l
4 j
Upon completion of radiological work involving surface
{
contamination, personnel will perform a complete whole body frisk.
j Frisking instruments will be provided at the boundary of each RCA i
and maintained by Health-Safety. Radiation Workers are trained on l
the use of frisking instruments during training; required for persons j
j permitted unescorted access to RCAs. All instruments will be l
selected based on the type of radiation or contamination being l
monitored within the particular area or the scope of work being performed. Friskers shall include a visual and audible alarm. The alarm set point will be established as low as possible taking into i
account the need to minimize the number of false alarms.
12.6.5 Personnel Decontamination
)
l Health-Safety shall be notified if contamination above the frisker l
alarm set point is detected on personnel skin or clothing as they i
exit from a RCA and when initial decontamination efforts fall to j
reduce the contamination below the alarm set point. Health-Safety shall assist with further decontamination efforts as necessary to i
reduce the exposure to a level as low as reasonably achievable, i
consistent with good health physics practice, before releasing the employee.
12.6.6 Contamination Control Enclosures l
Hoods, glove boxes, and containment tents are installed and j
utilized to control the spread of contamination and airb.orne radioactivity to levels as low as reasonably achievable. The type of
}
enclosure required will be determined by the individual operation l
being performed, evaluated by Health-Safety based on the (a)
{
contamination type and levels, (b) potential for generation of I
l i
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' FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY l'
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 12 - RADIA TION PROTECTION 1
airborne particulate,
- proximity of workers to the operation, and j
(d) previous operational history and exposure information.
i Contamination Control Enclosures shall be constructed of fire resistant materials including viewing ports and windows. Negative 3
differential pressure shall be maintained at a minimum of 0.25" of water when all openings are closed.
1 12.6.7 Action Umits ii
)
l Action limits are established for the purposes of maintaining proper j
contamination controls. When contamination above a specified action limit is detected, Health-Safety is responsible for conducting 4
of review of the situation or work in progress to determine l
corrective actions and directions for returning the area to normal i
levels. The action limits for each area are specified in section 3.6 l
i 12.7 Airborne Radioactivitv i
l 12.7.1 Ventilation Svstem I
j Airborne radioactivity in permanent Radiologically Controlled Areas -
j shall be controlled by use of an engineered ventilation system. The j
ventilation system shallinclude heat resistant High Efficiency l
Particulate Air (HEPA) filters with a minimum rated collection efficiency of 99.95% @ 0.3 micron DOP particulate. Effluents i{
from the ventilation system will pass through a minimum of a single stage HEPA filter. Air recirculated into a work area will pass through a minimum of a single stage HEPA filter.
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j Air c'apture devices shall be provided in the work area to remove airborne contaminants from the breathing zone of workers in the j
area. Minimum face velocities shall be determined for each such device and the performance monitored at regular intervals to ensure continued acceptable, performance.
A ventilation system will be provided for each Radiologically Controlled Area to maintain areas of higher contamination at a slight negative pressure to uncontrolled areas.
Page: 12-6 December 30,1996 Revision: 3
FRAMATOME COGEMA FUELS - L YNCHBURG MANUFACTURIN3 FACMJTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARYM - CHAPTER 12 - RADIA TION PROTECTION d
i 1
12.7.2 Air hmnlina Proaram To verify the effectiveness of the ventilation systems and j
contamination control enclosures or air capture devices,_ air samples in work areas at fixed or permanent locations wul be 4
collected and analyzed. Breathing zone air (BZA) samples will be
{
collected for the purposes of determining exposures as provided in Title 10, Code of Federal Regulations, Part 20 for employees j
performing radiological work. Collection of breathing zone samples shall be performed in accordance with Regulatory Guide 8.25, " Air Sampling in the Workplace". Use of BZA samples shall be dictated by the RWP for the assigned task (s). Evaluation of air sample.
results will be conducted to ensure compliance with applicable exposure limits and maintain appropriate cognizance of work area j
conditions. Work operations will be reviewed and evaluated when i
employees receive an exposure in excess of 4 DAC-Hrs for routine j
work or 6 DAC-Hrs for non-routine or infrequent work. The maximum evaluation period interval for routine work will be 1 shift j
(the largest period of continual work by one individual per day) and for non-routine work, the duration of the work operation. Fixed or static air sample results will not be evaluated on a DAC-Hrs basis, j
but will be used to ensure that the concentration in the work area 1
is consistent vsith BZA measurements. Average quarterly concentrations are calculated an'd reviewed for the purposes of j
verifying the continued effectiveness of the ventilation system.
l Effluent measurements are performed to determine the amount of radioactive material released to the environment. Effluent air j
measurements are evaluated to ensure compliance with exposure limity for the general public and environment as required in 10 CFR j
and 40 CFR.
i
~
12.7.3 Resniratorv Protection l
3 A~ supply of respirators, equipped with high efficiency filters, is j
maintained by Health-Safety for use when airborne concentrations are such that personnel may receive unacceptable levels of exposure from airborne radioactive material or other airborne l
hazards. Respiratory protection devices are available that provide protection from particulate, acid gases, ammonia, and other toxic substances. Self-contained breathing apparatus (SCBA) units area 1
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- FRAMATOME COGEMA FUELS - LYNCHBURD MANUFACTt/ RING FACillTY
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i.
USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTM - CHAPTER 12 - RADIA110N PROTECTION
[
l f
available for use in emergency situations. When respiratory j
protection devices are employed, the appropriate protection factors j
are specified in 10CFR2O are applied to the measured
~
concentration, yieldin'g a calculated exposure to airborne contaminants. Operations which are known to generate airborne i
contaminants include the removal and replacements of particulate j
air filters in the HVAC system and cleaning of systems contaminated with dry uranium powder. Other operations are evaluated for Respiratory Protection Device usage when the RWP l
for the work is developed. Personnel qualified to wear respiratory j
protection devices are trained in the proper methods of fit, fit j
testing, and removal. Proper fit is verified by a nuisance atmosphere during the training program. Maintenance of j
Respiratory Protection Devices is based on the manufacturer's j
recommendations and is performed by Health-Safety. Routine i
inspections include checks for operability and deterioration and replacement of outdated components. Outdated canisters may be used in training exercises, but will be marked to prevent use during i
a situation where protection is actually required. Air Sampling of the worker's breathing zone is always performed when respiratory protection devices are employed.
12.8 Radioloalcal Survelliance and Monitorino j
12.8.1 Routine Survevs To monitor the radiological conditions in work areas and ensure the continued effectiveness of the radiological control program, routine j
surveys will be conducted. Routine surveys shall include surface j
contamination and radiation levels in and adjacent to radiological
{
controlled areas (RQAs) and contaminated areas. When operations -
l are suspended for more than a few days, survey frequencies will be adjusted accordingly to match the extent of the corresponding operations or activity in the area. Surveys shall also be conducted -
j in uncontrolled areas to monitor the spread of contamination i
beyond established contamination control boundaries.
j 12.8.2 Instrumentation j
Survey techniques and instrument selection will be selected based on the isotopes and type of radiation being monitored. All
}'
Page: 12-8 December 30,1996 Revision: 3
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACit/ RIND FACillTY USNRC LICENSE SNM-1968, DOCKET 70-1201 PARTM - CHAPTER 12 - RADIA TION PROTECTION instruments will be calibrated at six month intervals except where a longer calibration period can be demonstrated as acceptable. Each instrument will be calibrated using standard radioactive sources which match the type of radiation being sensed by the instrument.
Laboratory instruments are characterized each day they are used and calibrated at a semiannual interval. Typical radiation and contamination survey instruments include GM tubes, gas sealed and flow proportional counters, scintillation detectors, alpha spectrometers, whole body frisking systems, and ion chambers.
Other instruments include portable flow measuring and calibration kits, positive displacement air pumps, air flow meters, combustible gas indicators, sound level meters, and toxic material detectors.
12.8.3 Biome =av Program Radio bioassay will be performed to determine whether personnel exposure assigned by airborne radioactivity measurements are accura.te. Techniques willincJude analysis of urine samples and lung or whole body counts. Where greater sensitivity is required, other types' of bloassay measurements or techniques maybe employed. Bioassay measurements will not be used to determine or assign routine personnel exposures, in the case of significant internal exposures, bioassay measurements may be used to supplement or separately determine the amount of internal exposure received.
12.8.4 Criticalltv Monitoring Svstem A criticality monitoring system shall be maintained in compliance with }he appropriate sections of 10 CFR 70. Response time for the system shall be,in accordance with Regulatory Guide 8.5,
" Criticality and Other Interior Evacuation Signals" dated March
~
1981. The criticality monitoring system will be functionally tested at least quarterly and detecter units calibrated annually.
12.8.5 Release for Unrestricted Handling (RFUH)
Equipment and areas previously contaminated above the limits specified in section 12.6 of this document shall be surveyed and released for unrestricted handling (RFUH) in accordance with
" Guidelines for Decontamination of Facilities and Equipment Prior to Page: 12-9 December 30,1996 Revision: 3
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FRAMA TOME COGEMA FUELS - L YNCHBURG M/NUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKEY 70-1201 PARTH - CHAPTER 12 - RADIA TION PROTECTION i
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Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material",'USNRC, August 1987, Exhibit A to section 1 of this document.
12.9 Renorts and Records 12.9.1 Records and reports pertaining to Health-Safety activities and i
employee exposures are maintained and stored in accordance with Regulatory Guide 8.7, "Occunational Radiation Exnosure Recordc Systems" Specific provisions for record retention and report 4
j submittal are addressed in the appropriate sections of 10 CFR.
Documentation of Health-Safety activities are performed by or.
i under the administrative control of the Health-Safety Organization.
Records pertaining to radiation and contamination surveys, radioactive material shipments, personnel exposure measurements and calculations, exposure reports, instrument calibratlons and maintenance, audits and inspections, and employee information files are included in the Health-Safety records system.
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5 FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURINJ FACIUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 13 - ENVIRONMENTAL SAFETY-RADIOLOGICAL l
j 13.1 General
~
A complete description of the environmental monitoring program used at the LMF is given in Chapter 5.0. Subject areas presented in Chapter 5 l
include:
i the location of sampling points, and monitoring and ba'ckground stations 1
the type of samples taken
~
the frequency of sample collection the minimum sample size the type of sample analysis the minimum detectable level Figure 13.1 (the same as Figure 5.1) is a site map showing the location of the different sampling locations and type of sample taken. Tables 13.1 and 13.2 are summaries of the data taken at each station for 1990 and 1991.
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I FRAMA TOME COGEMA RIELS - LYNCHBURG MANUFACTURIN3 FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 13 - ENVIRONMENTAL SAFE 1Y-RADIOLOGICAL 4
j i
l TABLE 13.1 1
ENVIRONMENTAL SAMPLE RESULTS CALENDAR YEAR 1991 I
t.-
4::.
< x 3
s..-.
.,3......
.-e..
r.n..
o s......
~
N NU$5 5.WO@!
- M M+
, 2 h:MDh
)
'M'..
4
[. Eh
' [W s
"gggrjY
[M f:
{
],
- f..
f
[i!
g3pgMt N
g j
tyg;I!$
fjQh Fjpgi ffG;_
Vfffw kJN ge:
yfh3 c
s5
> llLf y vgi Xfa
> lQit '
s $$9
?jis
@ !P N 1Mi ff?
(f9 g,ggy a
g g
A 2.7 3.6 1.2E-15 1.5E-14
- 0. 3 0.1 31.80 j
8 09 0.3 3 3E-15 3 GE-14 0.3 0.1 28.48 C
1.2E 15 1.7E-14 30.86 3
i i
D 1.1 2.4 i
E 68 12.0 2.4 4.7
(
)
F 0.9
- 2. 8 1.4E-16 6 SE-16 0.3 0.1 32.10 i
i A
46 7.8
]
I H
0.7 2.5 0.3 0.1 1
j 1
28 80 1.0 2.4 4
J 2.3 6,3 1.6 3.1 4
)
x 2.8 v.8 1.i i 4. 3 i
L 7.0 11.8 1.9 4.9 4
Station G value (soil) is the average value of all samples taken in Calendar Year 1991. There are 9 different sampling locations for Station G that are each j
sampled quarterly. The value indicated is the average of all 36 samples.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACiUTY USNRC UCENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 13 - ENVIRONMENTAL SAFETY-RADIOLOGICAL TABLE 13.2 ENVIRONMENTAL SAMPLE RESULTS CALENDAR YEAR 1990
' TYPES d.%f s.sys j[AM'$
!IA L. WATWA 4 asnanasary
);vgegy[,
iTLD3g!
UNTT3
[~.M:[
[M DPMut l1[ M M %
s / M'i-
- . ':Mi
.c.g fgn F yM.
s'jyg
? jy'g't l' y';;)
ygg:;;
- y gp; 9
g 9s I90 l Nk0 Ob?
?N$$
.g,
.g%
9ggi SA r..;
290 l Ii E80V A
3 A
6.7 13.8 1.3E 16 1.70 14 0.3 1.1 21.35 B
3.0 7.4 0.3 16 18.28 C
1 GE 16 4 OE-14 20.98 0
4.6 16.9 E
- 3. 8 7.0 4.B F
2.5 to 1
- 1. ?E-16 1.6E-14 0.3 1.4 18 45 g
O 21,6 21.4 H
4.5 17.0 0.3 1.9 1
23 66 1.7 6.2 i
J 2.0 36 1.6 48 i
K 2.0 6.3 1.3 4.1 t
12 0 11.3 36 16.6 Station G value (soll) is the average value of all samples taken in Calendar Year 1990. There are 9 different sampling locations for Station G that are each sampled quarterly. The value indicated is the average of all 36 samples.
,\\
i I
j 1
Page: 13-3 December 30,1996 Revision: 2
.=_
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURIN3 FACIUTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 13 - El/VIRONMENTAL SAFETY-RADIOLOGICAL i
FIGURE 13.1 ENVIRONMENTAL MONITORING
- 1 STATION DIAGRAM 1
i 3
g B
A fl.
i
\\
i I
i L _ "il l
K D
i A
j['
,H l
O 1
,1], l, 3
i L %<-- ---
f' gl
- G-a i
/
l F
ua.=""
. /'.__....._
wx j
i w.
.i i
!%~
A&K - Background Stations 1
Station Identification Sample Type A
j Soil, Air Vegetation, TLD B
Soil, Vegetation, TLD, Air C
Air, TLD D
Soil E
Water, Sediment F
Soil, Air Vegetation, TLD G
Soil H
Soil Vegetation I
Water, Sediment J
Water, Sediment K
Water, Sediment L
Water, Sediment Page: 13-4 December 30,1996 Revision: 2
i FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACnilMW3 FACillTY
^
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY 14.1 Administrative and Technical Procedures The ultimate responsibility for nuclear criticality safety at the LMF rests j
with the LMF Plant Manager. The Plant Wlanager has assigned this responsibility to the Safety Review Board Chairman (Manager, Safety and Licensing) who must approve all modifications or additions which involve hazardous as well as nuclear materials. The Safety Review Board Chairman will determine if Board review is necessary. Written procedures i
i approved by plant management shall be utilized for all operations involving SNM. Nuclear safety postings approved by Health-Safety shall be maintained specifying nuclear safety parameters that are subject to procedural controls. The training program conducted at the LMF (Sections l
}
2.5 ana 11.5) provide additional assurance that the criticality safety i
requirements are adhered to. Administrative controls for performing and approving criticality safety analyses are described in Section 4.1.
14.2 Preferred Annroach to Deslan The double contingency principle as defined in the American National l
Standard ANSI /ANS-8.1 shall be followed in establishing nuclear criticality safety for all equipment, systems, and operations. Where possible and practicable, reliance will be placed on equipment design in which dimensions (i.e., favorable geometry) are limited rather than on l
administrative controls. Where structural integrity is necessary to provide j
assurance for safety, the design and construction of the equipment will be made with due regard to abnormal loads, accidents, and deteriorations.
14.3 Basic Assumntions The basic nuclear criticality safety limitations discussed in Chapters 4,14, and 15 were developed assuming UO, with a maximum U-235 enrichment of 5.1 weight percent and a maximum pellet diameter of 0.4 inches.
Uranium pellets processed at the LMF can however range in diameter from l
1 0.3 to 0.6 inches, and may include UO, powder and pellet chips. All data -
given is valid for the most reactive heterogeneous geometry appropriate to the situation being considered. The U-235 enrichment may not exceed 5.10 wt. % with measurement uncertainties.
4
)
l Page: 14-1 December 30,1996 Revision: 4
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTt/ RING FACill1Y USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY i
14.3.1 Enrichment Evaluations and limits discussed in this chapter and in Chapters 4 and 15 were based on a maximum enrichment of 5.1 wt. % U-235.
With product and statistical variations, the enrichment cannot i
exceed 5.10 wt. % U-235.
i Figures 14.1 through 14.4 illustrate basic critical parameters (mass, volume, cylinder diameter and slab thickness) versus enrichment for optimum moderated and reflected UO,, rods in water as taken from DP-1014. These graphs demonstrate the maximum sizes for l
Individual units and only a small extrapolation is needed to obtain j
values up to 5.1 wt% U-235.
l l
14.3.2 Cladding Material Where appropriate, zircaloy type cladding material was input to the nuclear safety calculations. Stainless steel cladding provides an
]
additional conservatism due to its higher effective cross section for neutron capture. Other clad materials may be used provided there j
is not an adverse affect on nuclear safety specifications.
4 3
14.3.3 Calculated safe Unit
)
A calculated safe unit is defined in terms of nuclear safety as having a K, A O.87 for normal operating conditions and a K,1 0.95 under assumed accident conditions. The statistical and methodological limits of error will be considered when determining K,#,.
\\
14.3.4 Calculated Safe Arrav A calculated safe array is defined for nuclear safety purposes as having a K,10.87 for normal operating conditions and a K, <.
0.95 under assumed accident conditions. The K, values for these calculated safe arrays will also consider statistical and methodologicallimits of error.
Page: 14-2 December 30,1996 Revision: 4
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACellTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY I
14.3.5 Safetv Factors for Simnia Shnnan i
f For simple shapes, safe values are determined by application of the j
following limits:
s Mass:
45% of minimum critical reflected mass I
i Mass:
75% of minimum critical reflected mass when l
i double batching is not credible Volume:
75% of minimum critical reflected volume l
Diameter:
90% of minimum critical reflected cylinder diameter Thickness:
90% of minimum critical reflected slab thickness It is recognized that interaction between simple shapes must also be accounted for.
14.3.6 Nuclear Interaction Acceptable nuclear interaction between arrays is determined using calculations or as stipulated in Section 4.2.3.1. The limits are determined for each process as it was evaluated and its limits were set.
The interaction restriction given in 4.2.3.1 is a generally accepted practice for assuring nuclear isolation of subcritical accumulations under all conditions if a calculated value is not available.
14.4 Fixed Poisons Fixed neutron poisons are utilized as part of the nuclear criticality controls in the Pellet Storage Vault. Specifications in the Pellet Storage Vault layout are given in Section 4.2.4.2. The criticality safety analyses performed specifically on the Pellet Storage Vault are given in Section 15.2, 14.5 Structural Integritv Policv and Review Program LMF plant policy regarding engineering design review for structural integrity and safety margins is indicated in Sections 14.2 and 2.3.
I Page: 14-3 December 30,1996 Revision: 4
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACIUTY USNRC UCENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NUCLEAR CRITICAUTY SAFETY 14.6 Analvtical Methods and Their Validation Computer codes were used to calculate K.,, of individual units and of arrays. Computer codes and associated cross section sets are always benchmarked to demonstrate their validity for application at the LMF. The l
particular family of codes may change over time as improvements are realized in the calculational methods. The LMF's goal is to maintain its computing capability at state-of-the-art for nuclear criticality safety and to j
assure that all codes and cross sections are properly benchmarked and validated.
i 14.6.1 Phvsics Comouter Codes All the calculations in this evaluation were done using the CSAS2 l
module of the SCALE-3 computer code package (reference 1). The SCALE (Standardized Computer Analyses for Licensing Evaluation) code package was developed for the USNRC by the Oak Ridge I
National Laboratory. Codes within SCALE-3 used by the CSAS2 3
module for nuclear criticality safety analyses are the cross section 1
processing codes NITAWL and BONAMI, the one-dimensional trans-port code XSDRNPM (when needed for cell averaging), and the 4
three-dimensional Monte Carlo code KENO-IV. The 123GROUPGMTH cross section library was used for all cases. Any additional nuclear criticality safety evaluation requiring the calculation of the effective neutron multiplication (K-eff) will be made using a validated and benchmarked version of an industry
+
I accepted computer code package such as the SCALE computer code package. Safe units are those with a true k,,, below 0.87 under normal conditions and below 0.95 under accident conditions.
The{true k,,, may differ from the k,,, value from a properly modeled KENO calculation because of the calculational bias and statistical uncertainty. The calculated upper limit of true k,,,is the value from KENO plus two standard deviations (also from KENO) plus a 0.02 bias. It is this true k,,, which is compared with the 0.87 and 0.95 limits to assure that a unit is safe. The 0.02 bias has been deter-mined to be the maximum non-conservative delta K.,, of our SCALE calculations based upon benchmark calculations of critical experiments.
The codes and cross sections of SCALE have been well validated for criticality safety analyses. A limited number of computer cases Page: 14-4 December 30,1996 Revision: 4
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY 8
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTN - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY t
i j
were provided within the SCALE manual which represent the evaluation of actual critical experiments; these not only provide an initial basis for validity but also provide a means to verify that the codes are performing as planned on the computer system being used. In addition to these cases, a number of independent studies
)
~
l have been performed which demonstrate the reliability of the codes and cross sections in SCALE when they are used to calculate various classes of critical systems. See paragraph 4.2.3.3 for j
description of types and range of variables used in validating the code package. In addition to the above, FCF's nuclear criticality l
l safety group has an on-going validation program which documents l
the results from current calculational methods and compares them with benchmark critical experiments. These comparisons are the l
basis fo-the 0.02 bias which we currently add to all calculations.
l Our Nuclear Criticality Safety Benchmark Notebook along with j
numerous internal memos document these comparisons.
i 14.7 Snecial Controls j
b The fuel assembly processing area (i.e., assembly room) is that plant area 4
l where the loaded fuel rods are configured into a reactor type fuel i
assembly. This area is under moderation control and as a result, the j
following conditions apply:
)
No sprinkler systems are permitted in the assembly room.
1 j
1 gallon volume restriction on containers in the area to preclude l
any meaningful moderation in the event of a spill.
]-
Baffling / shielding of water piping in the area to prevent general j
area moderation in case of pipe rupture.
Simultaneous application of more than one fire hose is prohibited.
4
]
The fuel assembly storage area is not a moderation controlled area but the use of more than one fire hose in the area is still prohibited.
{
Also, the fuel assembly dust wrappers are arranged to permit free
]
drainage of water from within.
i l
i Page: 14-5 December 30,1996 Revision: 4 1
i
FRAMA TOME COGEMA FUELS - L YNCHBURG ALANUFACTURING FACILITY l
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICALITY SAFETY 4
' 14.8 Data Sources l
The following documents will be used as sources of applicable data:
~
1.
ANSl/ANS-8.1-1983, Nuclear Criticality Safety in Operations with l
Fissionable Materials Outside Reactors.
j i
2.
TID-7016, Rev. 2 (NUREG/CR-OO95), Nuclear Safety Guide.
3.
LA-10860-MS " Critical Dimensions of Systems Containing 23sU, 28'Pu,and 2330 -1986 Revision.
4.
LA-3366 (Rev.), Criticality Control in Operations with Fissile Material.
5.
ARH 600, Criticality Handbook, Volume 1, 2, and 3.
6.
DP-1014, Critical and Safe Masses and Dimensions of Lattices of U and UO Rods in Water.
2 14.9 References 1.
NUREG/CR-0200, ORNL/NUREG/CSD-2, " SCALE: A Modular Code System For Performing Standardized Computer Analyses For Licensing Evaluations," Prepared by the Staff of the Nuclear Engineering Application Department, Union Carbide Corporation.
l
\\
Page: 14-6 December 30,1996 Revision: 4
FRAMA TOME COGEMA RIELS - L YNCHBURG MANUFACTURIN3 FACLUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NUCLEAR CRITICAll1Y SAFETY l
FIGURE 14.1 l
CRITICAL VALUES FOR UO2-WATER MIXTURES (FULI. WATER REFLECT 1oN ASSUMED) 5.00 -
4.50 -
i 4.00 J 2n
'i' 3.50 -
o
,.. ::>J 3.co -
Es 2.50 -
N \\
2.00 -
4,30
\\
1.50 i
i 1.50 2.50 3.50 4.50 5.50 WOGHT PERCENT U-235 0
1.84 KG e 4.10%
l l
Page: 14-7 December 30,1996 Revision: 4
FRAMATOME COGEMA FUELS - LYNCHBUR3 MANUFACRIRING FACallTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY FIGURE 14.2 CRITICAL VALUES FOR UO2-WATER MIXTURES (FUL1. WATER RERICTION ASSUMED) 40.00 39.00 -
38.00 -
37.00 -
36.00 -
35.00 ;
34.00 -
33.00 -
m 32.00 -
g 31.00 -
]
30.00 -
29.00 -
+.4%'i 28.00 -
4.10 27.00 -
26.00 -
25.00 -
24.00 -
23.00 -1 N
22.00 21.00 -
I 20.00 i
i i
i i
i i
i i
i T-i 2.40 2.80 3.20 3.60 4.00 4.40 4.80 5.20 5.60 WOCHT PERCENT U-235 0
26.6 UTERS O 4.10*:
Page: 14-8 December 30,1996 Revision: 4
FRAMA TOME COGEMA FUELS - L YNCHBUR3 MANUFACTURIN3 FACIUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY FIGURE 14.3 1
1 l
CRITICAL VALUES FOR UO2-WATER MIXTURES l
(FULL WATER REF1.ECT10N ASSUMED)
=
40.00 39.00 -
38.00 -
37.00 -
36.00 -
35.00 -
34.00 -
3 33.00 -
3 32.00 -
5 31.00 -
30.00 -
g p
29.00 0
,.,, 28.00 6 ~ " 27.00 -
26.00 -
0 25.00 -
24.00 ]
23.00,
22.00 M d
21.00 20.00 I, 1.50 2.50 3.50 4.50 5.50 WOGHT PERCD(T U-235 0
24.9 CMS. O 4.10%
1 Page: 14-9 December 30,1996 Revision: 4
~
FRAMA TOME COGEMA FUELS - LYNCHBUR3 MANUFACTURIND FACKITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY I
FIGURE 14.4 CRITICAL VALUES FOR UO2-WATER MIXTURES (FULL WATER REF1ECTloN ASSUMED) 20.00 19.00 -
18.00 -
17.00 -
m 3
[
1s.00 -
0 5
15.00 -
14.00 -
5*
13.00 -
12.00 H
,30 11.00 -i
\\
10.00 i
i 1.50 2.50 3.5o 4.50 5.5o WElGHT PERCENT U-235 0
11.5 CMS. O 4.10%
Page: 14-10 December 30,1996 Revision: 4 J
i l
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURN3 FACRJTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS 4
l l
15.1 Procean Stens and Flownheets i
l Section 10.1 is a brief description of the LMF plant layout and operations.
l This and the process flowsheet shown on Figure 15.1 describes'the f
process and material flow at the LMF.
j 15.2 Safety Analvnin and Safety Featuren of Each Stan l
)
This section will describe the safety features and analyses for each process step. A detailed discussion of both nuclear and radiological safety will be included. Each process step was analyzed for all degrees of moderation that might be created by fire fighting equipment. Where moderation must
{
be restricted it is done by controlling fire fighting methods and limiting the i
available moderating materiais at the process step. All processing steps have also been evaluated at the maximum allowable enrichment of 5.1 wt% U-235. When limits with regards to enrichment are encountered, l
restrictions are placed on the handling pf fuel with different enrichments.
j An example of this is allowing only a finite 25" wide 4" thick slab with 12" horizontal spacing restriction on unciad fuel with enrichments over 4.1 l
wt%. The safety analysis will utilize the same basic designations as i
shown on Figure 10.1.
l 15.2.1 Pellet Recelot i
i Nuclear Safetv l
1 l
Fuel pellets are received in NRC approved shipping containers in the " pellet receiving" area. All packages received containing j
special nuclear material are assumed to be packaged iri accordance l
with applicable regulations and license conditions. As a result, no special controls are required for nuclear safety of this array.
The fuel pellets are received in cardboard boxes. The 91/8" X 8 1/4" X 41/4" outside dimensioned cardboard boxes have a maximum fuel height of 4". During the initial unloading of the
}
shipping containers, the boxes are stacked two high (8 inch slab).
l The 8 inch slab occurs because each container contains 6 boxes.
i The boxes are stacked two high (8 loch slab) three boxes wide and they are transported to the conveyor in this arrangement. They are placed one high (4 inch slab) unto the conveyor. Only one i
3 i
1 l
Page: 15-1 December 30,1996 Revision: 5 I
J
i FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACRJTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
container is unloaded at a time so that only six boxes of material exceeds the four inch criteria at any one time.
i The local safety rules, training, audits, and pre-operational evaluat;ons ensure that the four inch slab criteria is maintained. All j
of which has proven to be effective.
i The pellets are arranged in layers that are separated by 0.017" l
thick stainless steel trays. The whole stack of pellets and trays is i
sealed in a 0.006" thick polyethylene wrapper which is then placed j
into a cardboard box. Layers of fuel pellets in a box are no longer l
separated by sheets of polyethylene.
1 Radiological Safety i
There are no major concerns regarding radiological safety in this area. The pellets are packaged in plastic inside cardboard boxes and do not present an airborne radioactivity problem. External radiation is low due to the nature of the material and therefore
. require no special controls or shielding.
15.2.2 Pellet Receiving Convevor Nuclear Safety The fuel pellets as received are transferred from the Pellet Receipt Area to the Pellet Vault via a conveyor. This 20" wide conveyor is located along the east wall of the South Bay. The nuclear safety of the conveyor is maintained by limiting the slab thickness and width along the length of the conveyor. Allowable fuel slab thickness will be fixed at a maximum of 4" and will no longer be a function of the Um enrichment. A 4" thick by 25" wide slab has been established as the limit for all fuel enrichments up through 5.1 wt% Um on the conveyor. The length of the conveyor was assumed to be infinite.
This slab thickness and width applies to fuel pellets being received and also to boxes of scrap pellets that are being shipped away l
from the LMF.
To prevent the four inch slab from being exceeded, there !s a physical barrier that would prevent anything that exceeds four inches in height from entering the pellet loading room. By design, Page: 15-2 December 30,1996 Revision: 5 ai
~--y
... ~
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i FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168a DOCKET 70-1201 1
l PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
l the conveyor limits the slab width to 20 inches. If other fuel accumulations are required to be in the area, the local safety rules, training, audits, and pre-operational evaluations ensure that the spacing criteria is maintained. All of which has proven _to be effective.
i i
The boxes are wrapped securely in plastic to help prevent the i
i boxes from breaking open and spilling the pellets.
)
I Scrap pellets being returned to the vendor are packaged the 91/8" X 81/4" X 41/4" outside dimensioned cardboard boxes and have l
a maximum fuel height of 4". The corrugated trays are not normal-ly used and typically only four boxes are shipped per container.
~
l The boxes are transported from the pellet loading room to the truck j
via the conveyor. For scrap shipments, the four inch slab is not exceeded.
Radioloolcal Safety The pellet boxes on the pellet receiving conveyor are not opened prior to entry to the pellet storage vault. As a result, there is no potential for airborne contamination and no need for any special Health-Safety controls. Since these are low enriched, unirradiated UO, pellets, special controls or shielding would not be necessary for control of external radiation.
15.2.2.1 Convevor calculations The conveyor was evaluated as a 4" thick x 25" i
wide x infinitely long slab with 5.1 wt% Um fuel.
The stainless steel trays which are a mild neutron poison were omitted as a conservatism for pellets being received and to account for the fact that no stainless steel trays are in the outgoing scrap fuel -
pellet boxes. The use of a 25" wide slab rather than 20" and assuming 100% theoretical density fuel pellets also added conservatism to the evalua-tion. In addition, the conveyor structure was also omitted from the calculations.
Page: 15-3 December 30,1996 Revision: 5
},
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACIU1Y USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS
~
i l
15.2.2.2 Calculational Results l
i CASE #1:
The computed km for a single 4" high j
and 25" wide infinitely long slab.having dry cardboard boxes filled with 5.1 wt% fuel l
pellets was 0.137 i O.002.
CASE #2:
Km for a double batched 8" high and 25" wide stack of dry cardboard boxes of 5.1 I
wt% pellets which is an overloaded accident l
condition was 0.211 f_ O.002.
CASE #3:
The conveyor is about 3 feet off the floor
{
but in the unlikely event that it should i
become flooded so that the 4" thick x 25" wide slab is optimally moderated and fully l
reflected, the k,,,,, is 0.919 i.006.
CASE #4 The infinitely long 4" thick x 25" wide slab i
of 5.1 wt% fuel can be located on a 12" l
thick concrete floor and the km will be 0.801 S. 006. If the slab is setting on the i
concrete floor and is then fully flooded and
{
reflected on the top and sides with water i
the k,,,,, will increase to 0.971 5. 006 i
which is above the allowable 0.95.
l Therefore; the 4" x 25" slab must be kept at i
least 12" above any concrete floor because of the increased reflection properties of con-j crete.
I 15.2.3 Fuel Pellet Storage Vault Nuclear Safety 1
Nuclear safety of the pellet storage vault is maintained by utilizing the safe geometric slab combined with separation distances and limited use of neutron poisons. The vault shelves have a barrier to prevent boxes of pellets from being stacked. In order to stack the boxes causing the 4 inch slab to be exceeded the worker would have to make a physical effort to stack one box on top of the Page: 15-4 December 30,1996 Revision: 5
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS other. The boxes remains securely wrapped in plastic during storage to help prevent the boxes from breaking open and spilling the pellets.
]
l The 25 inch width slab is maintained by the design of the conveyor, the vault shelves and the scissor transport cart all of which are less than 25 inches in width. The back of the hoods and 2
the down draft table is used for the 25 inch width allowing a 12 i
inch separation of " unusable" front space to maintain the spacing criteria.
j The local safety rules, training, audits, and pre-operational evaluations ensure that the four inch slab criteria, the finite slab, and the spacing criteria, is maintained. All of which has proven to be effective.
i Enrichment verification shall be determined by nuclear materials i
accountability acceptance of shipper's values or determined in house by allowable FNMC procedures.
1 The vault area consists of 3 " cubicles" that are used for storage of pellet boxes. A cross sectional view of one of the cubicles is j
shown in Figure 15.2. The figure shows the spacing between i
shelves, tiers of shelves, and the relative locations of the "Boral" j
poison material. A 20" wide conveyor for transporting fuel into or i
out of the vault is located 67 %" in front of the cubicles. The i
vault was constructed with high strength steel and designed for the i
same weight capacity that is licensed to date. Each shelf is rated i
for 1500 pound capacity and the maximum weight at full capacity j
is only 1071 pounds. It is not subject to any environmental ele-j ments and degradation of its integrity is not suspected. A l
summary of the dimensions of the storage vault is given below.
}
15.2.3.1 Vault Snecifications l.
Storage Racks (See Figure 15.2)
- No. of shelves / rack 5
- Shelf width 18"
- Shelf length 19'-0"
- Minimum horizontal 9'- 1 1 "
Page: 15-5 December 30,1996 Revision: 5
j i
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS i
spacing between shelves i
through air
- Minimum horizontal spacing 8"
between shelves through concrete j
- Minimum vertical separation 16" between shelves
- One side of the concrete wall is covered with B C poison plate. Four upper 4
l shelves on each tier of shelves are also
]
covered with B.C plate.
j ll. Poison j
- Material (Boral) 35 Wt.% B C in Al 4
- Plate thickness, In.
O.25"
- Core thickness, In.
O.168" i
- Clad thickness, In.
0.041" j
on each side of core l
- Core density, g/cc 2.46 111. Fuel J
- Material UO2
- Pellet Diameter, In.
O.4" (maximum O.D.)
- Storage Box (Cardboard) 9.125"x8.25"x4.25"
- Thickness, In.
4" (maximum)
IV. Moderation i
The cardboard storage boxes with a single polyethylene wrapper around the fuel inside the box is the normal storage condition. New fuel has stainless steel trays that separates each layer of pellets in the box. Scrap pellets-are in a single polyethylene bag inside the cardboard box.
The storage of UO pellets up to 0.4" diameter in the 2
vault will not present any criticality problems for a 4" or less slab thickness and for any enrichment below 5.1 wt% Um. This is with any amount of fuel and Page: 15-6 December 30,1996 Revision: 5
1 i
i' FRAMATOME COGEMA FUELS - LYNCHBURD MANUFACTURING FACIUTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
i moderator (water) in the boxes. Each tier of shelves can j
have boxes with different Um enrichments as long as the enrichment is below 5.1 wt%.
2 The following are conditions for justifications of the overall nuclear safety of the vault.
l 1.
The specified thicknesses of 4" maximum was j
shown to be safe for fuel enrichments up to i
5.1 wt%.
4 i
2.
The SNM array will also be safe for the unlikely i
accident of total flooding when the shelves are l
separated by a minimum of 12" vertical separation l
with 4.0" slabs of 5.1 wt% fuel.
1 2
3.
In the event of fire or an accident in which water is sprayed or sprinkled into the vault, the water i
should run down to the floor and flow out of the j
vault through door openings. Therefore, it is l
highly improbable that the voids in.the boxes will be completely filled with water and the space j
between the boxes have some amount of j
interspersed water. Even if this highly improbable l
accident situation occurs, the K.,,#, of the vault is less than 0.95 (with 2 sigma and 0.02 bias added) for all degrees of interspersed moderation and fully flooded boxes.
i 4.
The vault shelves can be inadvertently double loaded ( 8" thick slab) as long as there is no moderation present above the amount normally found in the boxes or vault. Double batching is not allowed and is considered an accident condition..
15.2.3.2 Vault Calculational Procedure The pellet storage vault and pellet handling area was calculated using the SCAL _E-3 computer code package using CSAS2. An infinite one dimensional planar array of Page: 15-7 December 30,1996 Revision: 5
i FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACKITY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
cubicles of the vault were modeled in KENO-IV showing
[
the 18" wide, 4" thick,19' long shelves as homogenized mixtures to easily account for various amounts of fuel and moderation in the boxes. The conveyor extending into the vault from the pellet receiving area and located j
67 %" in front of the storage shelves was modeled to l
have a 20" wide and 4" thick slab of fuel pellets on it.
All structural material in the conveyor was omitted. The "Boral" poison material was taken into account in the model but no shelf structural material was shown. The I
8" thick concrete walls,12" thick floors, and 8" thick
)
ceiling were assumed to be " Oak Ridge" concrete which tends to give higher or conservative k-effectives than
{
other types of concrete. The pellet processing table and l
rod loading table are located in the vault area but are separated from the storage section of the vault by an 8" l
thick concrete wall and a spacing of several feet. The pellet handling table and rod loading table were not i
modeled into the vault storage model because of the j
isolation created by the 8" thick concrete wall and distance. The pellet processing and rod loading tables l
were calculated using a finite 4" thick x 25" wide slab l
model as used for the fuel pellet conveyor.
i 15.2.3.3 Vault calculational Results CASE #1 The vault was calculated for normal con-
{
ditions. There was no water in the card-board boxes and no interspersed moderation f
)
around the ' boxes. The volume fraction of i
fuel in the boxes relative to amount of void was the normal amount for new fuel pellets j
that were stacked in orderly layers. This j
gives a higher k-effective for a dry box than-i using the pitch for optimum moderation.
I The stainless steel trays were replaced with j
void and the cardboard box was replaced j
with polyethylene for conservatism. The i
calculated KENO-IV km, was 0.423 5.
004.
I i
l Page: 15-8 December 30,1996 Revision: 5 l
1
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l FRAMA TOME COGEMA FUELS - LYNCHBURD MANUFACTURING FACIUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
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\\
4 i
CASE #2 The vault was calculated assuming normal conditions for the boxes as in case 1 but each shelf was assumed to have two layers,
of boxes (double loaded). The calculated K,
for tnis accident condition was 0.564 5. 004. There was no moderation in 4
j the boxes and no interspersed moderation around the boxes, i
l CASE #3 The boxes were assumed to again be in the i
normal conditions as in cases 1 and 2. The shelves were assumed to be collapsed to allow for all five tiers of fuel to be on top of l
each other and setting on the concrete floor.
The computed k, for this accident condition was 0.583 i.004.
CASE #4 With the boxes in the normal conditions and shelves intact, the Boral poison was removed from each shelf and from the walls between cubicles and replaced with void.
l The calculated k,. for this highly j
improbable condition was 0.584 i.005.
CASE #5 A sensitivity study was made with various degrees of interspersed moderation sur-rounding the fuel boxes. The interspersed moderation was varied from 0% up to a i
fully flooded condition of 100% interspersed i
moderation. It was assumed the fuel boxes
{
were filled with fuel / water at optimum moderation. The interspersed moderation j
was then varied from 0% to 100%. The results are tabulated in Table 1.and plotted in Figure 15.3. Because the boxes were assumed to always have optimum moderation, the computed k-effectives tended to be fairly constant for all degrees of interspersed moderation.
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70 '1201 PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS i
4 Table 1.
K-effective for Varying Amounts of Interspersed i
Moderation in the Fuel Pellet Storage Vault.
Percent i
Interspersed Moderation K-effective i 2 sigma O
0.886 i.006 j
1 0.888 i.005 l
2 0.882 i.006 3
0.878 i.006 4
0.879
.005 5
0.886 i.005 l
6 0.886 i.006 7
0.889 i.006 8
0.880 i.006 2
10 0.886 i.005 15 0.875 i.006 1
20 0.874 i.006 i
25 0.865
.006 i
50 0.872 i.006
]
100 0.883
.007 1
1 l
CASE #6 The pellet handling table, pellet leading i
table, and vault transfer cart will use the
]
same finite slab thickness of 4" thick and 25" wide as determined to be safe for the conveyor into the vault. The computed k-effective for a single 4" high and 25" wide i
slab with dry cardboard boxes filled with 5.1 wt% pellets was 0.137 i O.002. The finite slabs must also be separated from any other slab by a 12" minimum horizontal spacing. The slabs must also be at least 12" above any concrete floor. The concrete floor is a better reflector than water and will cause the finite slab to exceed 0.95 if it is setting on the floor and is then flooded with water. The finite 4" x 25" slab of 5.1 wt%
fuel can be located on the concrete floor and the km, will be 0.8011.006. If the Page: 15-10 December 30,1996 Revision: 5
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACLUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS i
slab is setting on the concrete floor and is I
then fully flooded and reflected on the top and sides with water the km will increase i
to 0.971 S. 006 which is above_the allow-
]
able 0.95. Therefore, the slab must be l
maintained at least 12" above any concrete
{
floor or slab.
CASE #7 Fuel pellets may be stored in areas other than inside the vault while still in the cardboard shipping boxes. Fuel pellets at j
4.1 wt% enrichment or lower can be stored j
in a 4" thick infinite slab that has a k,,,.
of 0.921 f_.006 when fully flooded and reflected. Fuel pellets with enrichments between 4.1 wt% and 5.1 wt% must be stored in a 4" maximum thick slab that is no more than 25" wide as calculated in the case for the conveyor into the vault from the pellet receiving area. The finite slab must be separated from any nher slab or fuel accumulation by a mini.num of 12" horizontally. All enrichment pellets must be stored at least 12" above any concrete floor i
or slab.
Radioloolcal Safetv The pellets are still in their cardboard boxes while being stored j
in the vault and therefore do not present an airborne contamination problem. The pellet boxes are opened upon initial receipt for sampling purposes. This operation is performed adjacent to the storage vault while the boxes are still on the conveyor after they enter the controlled area. The sampling station is equipped with an air pickup device to keep the airborne contamination levels below authorized limits.
Appropriate static and lapel air sampling devices as well as routine contamination surveys provide the monitoring necessary for contamination and personnel exposure control.
Page: 15-11 December 30,1996 Revision: 5
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3 FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS
)
i The pellet loading area is located within the controlled area due j
to the handling of unciad SNM. As with the pellet sampling i
station, the pellet loading aree utilizes appropriate air pickup devices to control airborne contamination. Contamination surveys and air sampling devices are used to verify the effectiveness of our air captura devices.
i 15.2.4 Fuel Rod Processing and Storage Nucienr Safetv l
l After the pellets are loaded into the fuel rod cladding, the rods are processed at various stations in the Fuel Rod Fabrication l
Area. The nuclear safety requirements for the safe handling and l
storage of loaded fuel rods are:
1.
Individual accumulations are limited to 4.0 inch thick slabs of fuel rods with enrichments less than or equal to i
5.1 wt% Um. Slabs can be infinite in size for dimensions other than thickness.
i 2.
One fuel accumulation shall not be horizontally positioned above or below another fuel accumulation of j
up to a 4" thick slab.
l 3.
One fuel accumulation up to a maximum of 4" thick slab can be vertically positioned higher or lower than another
{
fuel accumulation up to a 4" thick slab with no minimum horizontal spacing requirement other than the require-l
\\
ments of number 2 above. Analysis has shown that with full water reflection of the two non-coplanar optimum j
moderated slabs, there is no gain in r>sactivity by off-i setting the two slabs of fuel rods. At lower degrees of f
reflecting moderation, the k-eff is seen to increase slightly with the off-setting of the optimum moderated 1
slabs but the calculated k-eff is small enough to allow for l
the slight increase in k-eff.
i 4.
The 4" maximum thick slabs must be maintained at least 12" above the concrete floor. Reflector materials that provide better reflection than concrete such as beryllium i
Page: 15-12 December 30,1996 Revision: 5 I
J FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURIN3 FACILITY l'
USNRC LICENSE SNM-1168, DOCKET 70-1201
{
PARTM - CHAPTER 15 - PROCESS DESCRIP110N AND SAFETY ANAL YSIS and lead will not be allowed in any s!gnificant quantitles j
at LMF in the fuel processing areas.
l i
l To maintain the four inch slab, all fuel rod transfer carts have 4 4
inch channels with side rails to contain the fuel. Fuel rod storage shelves are labeled to instruct workers that fuel cannot be above or below another accumulation.
The local safety rules, training, audits, and pre-operational evaluations ensure that the four inch slab criteria and spacing criteria is maintained. All of which has proven to be effective.
15.2.4.1 Fuel Rod Proceaning calculations Nuclear criticality safety for fuel rod processing has been evaluated using the SCALE computer code package with the CSAS2 module. Calculations were made for 5.1 wt% Um zircalloy cladded fuel rods in a 4" thick infinite slab configuration. Two different size fuel rods were evaluated at several different spacings between rods while the rods were fully flooded and fully reflected.
)
This variation of pin pitches for the rods assured that the
]
optimum moderated situation was accounted for in the evaluation. Both fuel rod types k,,,,,, did not exceed 0.95 at any degree of moderation.
15.2.4.2 Fuel Rod Processing Calculational Results CASE #1 B&W's standard MKB fuel rods and MKBW fuel t
rods were evaluated in a 4" thick infinite slab.
4 The MKB rod has a 0.370" O.D. fuel pellet in a 0.377" 1.D. and 0.430" O.D. zircalloy cladding.
The MKBW rod has 0.3195" O.D. fuel pellets inside a 0.326" 1.D. and 0.374" O.D. zircalloy cladding. These two rods are typical of rods in 15x15 and 17x17 array fuel assemblies and they represent the range of rods being fabricated at the LMF. The two type rods were evaluated with l
varying distances between the rods. The distanc-es between rods were chosen to give a particular volume fraction of Nel in the triangular pitch cell.
Page: 15-13 December 30,1996 Revision: 5
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACIUTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS i
Table 2 gives the k-effectives for the various pitch rods when placed in a 4" thick infinite slab in KENO-IV. Figure 15.4 is a plot of the results.
Table 2.
K-Effective Calculated For 4" Thick Infinite Slab of MKB or MKBW Fuel Rods and Water at Various Volume Frac-tions of Fuel.
Volume MKBW MKB Fraction 0.3195" 0.370" Fuel Pellets Pellets 0.20 0.914 i.005 0.911 i.006 0.25 0.921 i.005 0.919 i.006 0.30 0.912 i.005 0.909 i.006 0.40 0.861 i.006 0.859 i.005 0.50 0.788 i.005 0.794 i.005 0.6618 0. 67 7 i. 0 0 5* * ------------
0.677 i.005**
0.6715
- Volume Fractions For Closest Possible Packing In A Triangular Pitch Array.
Radioloalcal Safety The Fuel Rod Processino rnd Storage area is a part of the uncontrolled area at th
&F. As a result, this area is under l
' clean area limits for contamination control (Section 3.2.6). The SNM that is processed in this area is encased in cladding and does not yield a meaningful potential for contamination problems. The loaded fuel rods do have a very small pressurization hole in one of the end caps, but does not present a contamination concern. This hole is welded shut after the rods are dried and pressurized.
In addition, air effluents to this area fall under the restrictions given in Section 3.2.2.2. Note that there is an exception at the end weld operation area and the end cap removal station to allow up to > 50% of the uncontrolled area MPC. Engineering Page: 15-14 December 30,1996 Revision: 5
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACIUTY USNRC UCENSE SNM.-1168, DOCKET 70-1201 1
PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS efforts to keep the air effluent at these stations under 10% of j
environmental MPC was not successful. Also, contamination surveys do not indicate a potential threat of contamination spread.
l i
l 15.2.5 Fuel Amenmbly Procanning i
Nuciaar Safetv i
1
+
j Fuel Bundle Assembly is shown as Area #11 on Figure 10.1.
l The assembly operation area is where fuel rods are loaded into j
the final configuration for a fuel assembly under moderation control conditions. The assembly area is a completely clos 3d room of concrete block construction with a metal deck roof l
topped by poured concrete. There is no sprinkler system in the assembly room. The use of cleaning materials is essential to fuel assembly operations. The restriction of twenty, one gallon i
containers effectively prevents the contents of the room from l
meaningful moderation in the event of a spill. The piping is protected from accidental rupture. The joints become the areas j
most susceptible to failure. Baffling of the joints to control the j
direction of water flow provides assurance that the assemblies j
in the fabrication process cannot become moderated. Some amounts of polyethylene type material may be in the area for dust protection of equipment, or as preassembly protective i
wrappers on components. Precluding the polyethylene from in-l terspersion among fuel rods other than when in the 4.0" thick
}
slab, completes the controls against moderation.
The nuclear safety of fuel assembly processing for four l
assemblies in the room spaced at 48" edge-to-edge spacing is determined using the SCALE-3 CSAS2 module. Assembly of i
fuel assemblies at distances of 38" center-to-center was also l
confirmed. The single Connecticut Yankee assembly table i
located beside the fuel assembly storage array is safe with the same spacing from any other fuel accumulation.
The bundle assembly tables weigh about 8000 pounds a piece i
preventing any movement that would violate the spacing criteria.
i h
l Page: 15-15 December 30,1996 Revision: 5 1
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY
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USNRC LICENSE SNM-1168 DOCKET 70-1201 PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS l
The local safety rules, training, audits, and pre-operational evaluations ensure that the moderation control and the spacing j
criteria are rnaintained. All of which has proven to be effective.
15.2.5.1 A=mambiv Procamming Calculaticas l
The assembly fabrication tables were modeled in detail to j
show the 17 X 17 array of 0.3195" diarneter UO, fuel j
pellets clad with zircalloy and spaced on a 0.496" pitch in a fool assembly. Twenty-four guide tubes and one instrurnent cell were modeled leaving a total of 264 fuel rods in the assembly. The MKB 15 X 15 array of 0.370" diameter UO fuel pellets clad with zircalloy and spaced 2
on a 0.568" pitch with 16 guide tubes and 1 instrument j
cell was also evaluated at selected points. The assembly was setting 5.5" above an 18.5" high by 30.75" wide granite table. Fixtures and grids holding the fuel assembly were ignored for conservatism. The granite i
table was assumed to be the same length as the 144 inches long fuel rods. The granite table was setting on a 12" thick concrete floor. The fabricated assembly was reflected on one end to conservatively account for a possible full loading of 264 fuel rods being held on the console table a minimum of 12" away ready for insertion into a fuel assembly. The fabricated assembly was reflected on each side to simulate an infinite one dimensional planar array of fuel assemblies and console tables being located 48" edge-to-edge apart. A 38" center-to-center spacing between assemblies was also evaluated. All degrees of interspersed moderation from 0% to fully flooded were computed. For all cases of interspersed moderation, the array of fuel rods were assumed to contain the same density water within the array of fuel rods as the interspersed moderation surrounding the assembly. The 0% moderator case assumed no water inside or outside of the fuel assembly.
15.2.5.2 Fuel Assembiv Processing Calculational Results CASE #1 The MKBW fuel assembly was assumed to be fully assembled and setting above the granite table.
Page: 15-16 December 30,1996 Revision: 5
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l' USNRC LICENSE SNM-1168, DOCKET 70-1201 1
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l l
The assembly was positioned to be either 48" edge-to-edge from another assembly on either side or it was assumed to be only 38" center-to-center from assemblies on either side. The interspersed i
moderation was varied from 0% (dry) to 100%
(fully flooded). The MKB assembly was calculated for 100% moderation which is the limiting condi-l tion. The enrichment was 5.1 wt% Um for all cases. The density of the fuel in the assemblies l
was assumed to conservatively be 100% of J
theoreticalinstead of the 94% to 96% of l
theoretical density typically found in the fuel j
assemblies. Table 3 gives the results of the i
KENO-IV calculations while Figure 15.5 is a plot of i
the results. The fully moderated situation is the l
worst and with 2 sigma plus 0.02 bias added, the j
maximum k, for the MKB assembly is 0.966 l
which exceeds the allowable 0.95 k,.. for acci-t i
dent conditions. This stresses the importance of fabricating fuel assemblies in a moderation i
controlled environment. The 38" center-to-center
)
spacing gives only a slightly higher K,#, at limit-ing conditions than the 48" edge-to-edge. At the
]
worst case conditions of full moderation the k,
, will be about the same because there will j
be isolation between two fuel assemblies at both j
spacings. The limiting fully flooded condition is a j
highly improbable event because it would require I
the assembly room to be flooded to at least 40" j
deep with water.
i 2
i t
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACKITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS 1
j Table 3.
K-Effective for MKBW and MKB Fuel Assemblies on the Fuel Assembly Fabrication Table with Various Degrees of Interspersed Moderat' ion.
Percent MKBW MKBW MKB MKB Interspersed 48" ETE 38" CTC 48" ETE 38" CTC Moderation 0
.116 i.002 2
.306
.003 4
.342 i.004 6
.343 1.004 8
.348 i.004 10
.358 i.004
.434 i.004 25
.487 i.005
.492 i.004 50
.672 i.005
.676 i.005 l
75
.822 i.006
.823 i.005 100
.935 i.005
.939 i.006
.940
.005
.937 i j
.006 CASE #2 The fuel assemblies are assembled in a room that 4
I has four fuel assembly stations located side by side. The situation can occur where a finished assembly must pass over a partially finished assembly on its way to the transfer cart in the center of the room. The case of two assemblies passing next to each other was calculated to assure safety for this transient condition. The i
calculation assumed that the two assemblies were arranged one above the other and in contact with each other. The Km, for two assemblies being adjacent is 0.165.i. 002 for dry conditions. This transfer would never occur if there was any significant degree of interspersed moderation present in the room.
l Page: 15-18 December 30,1996 Revision: 5
l FRAMATOME COGEMA FUELS - LYNCHBURG AGANUFACTURING FACRJTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTM - CHAPTER 16 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
Rarilologtt al Safety i
The Fuel Bundle Assembly area is an uncontrolled area that processes fully clad SNM. Those radiological controls used for l
clean areas are appropriate for this area. There have been no contamination or personnel exposure problems in this area.
l 15.2.6 Fual Annambly Storage Nucinar Safetv i
.)
The fuel assembly storage area is identified as Area #12 in l
Figure 10.1. The fuel assembly storage area is the primary i~
storage location for completed fuel assemblies prior to
{
shipment to the reactor site. Fuel assemblies are stored i
vertically in a 10 by 20 array of storage positions. The array has 21" center-to-center spacing in the direction of the 10 positions and has 38" center-to-center spacing for the other direction. There are a total of 200 storage positions. The restrictions that apply to the fuel assembly storage location are given below.
The finished fuel assemblies are stored without any control rods or other types of inserts.
The assemblies are wrapped with 0.006" thick polyethylene dust covers that are fixed to assure that the stored assemblies are free draining to prevent any accumulation of water.
\\
The fuel assembly storage array does not have any steam sources located near it.
There are no sprinklers located near the storage array. -
As a further conservatism, the use of hoses to fight any possible fire in the storage array is prohibited unless prior authorization has been obtained from a knowledgeable i
management representative (i.e., the plant manager, the emergency officer, the Health Safety Officer) of the plant emergency organization.
Page: 15-19 December 30,1996 Revision: 5
FRAMATOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70- 1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS If the use of a fire hose is considered necessary, no more than one line will be applied to the storage array at any given time.
Restrictions on the use of hose lines shall be prominently posted at the boundaries of the storage array such that they are visible from all fire fighting approach routes.
The spacing requirements are maintained by the design of the fuel assembly racks. Each base plate is nine inches apart which results in a 21 inch center to center spac'ing. For > 4.1 wt% enrichment, there is a dedicated storage rack modified by the removal of every other base plate resulting in a 42 inch center to center spacing. The fuel assembly rack was designed for a load capacity greater than the actual weight encountered.
The local safety rules, training, audits, and pre-operational evaluations ensure that the spacing criteria is maintained.
The basic parameters for two of the typical square lattice UO, fuel assembly types that will be stored in the storage array are as follows:
Basic Assemblies MKBW MKB Fuel Rod Array Size 17 x 17 15 x 15 Fuel Rods per Assembly 264 205
\\
Fuel Rod Pitch (inches) 0.496 0.568 Guide Tubes in Assembly 24 16 Instrument cells 1
1 Fuel Pellet O.D. (inches) 0.3195 0.370 Active Fuel Length 12 Feet 12 Feet Fuel Cladding Material Zircalloy-2 Zircalloy-2 Page: 15-20 December 30,1996 Revision: 5
3 FRAMA TOME COGEntA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS k
l Cladding 1.D. (inches) 0.326 0.377
)
, Cladding O.D. (inches) 0.374 0.430 1
i l
Other fuel assembly designs different than the ones listed above may i
be processed at LMF. These other assemblies must meet appropriate i
nuclear safety criteria specified in the license.
i 15.2.6.1 Fuel Assemhiv Storage calculat!nns The fuel assembly storage area was calculated using the SCALE-3 computer code package with the CSAS2 module. Both the 17 x 17 and 15 x 15 fuel assemblies were modeled in detail to show an array of fuel pins with cladding. The instrument tube and control rod guide tubes were shown for the infinite array calculations but were omitted for the finite array calculations that used homogeneous fuel assemblies. A 144" active fuel length was used for both fuel types. All fuel assembly grids and end fittings were omitted for conservatism. All structural material in the storage racks was also omitted for additional conservatism. The fuel assembly was located 6" above a 12" thick " Oak Ridge" concrete floor.
A 120" region of interspersed moderation was located above the assembly. The assembly was located in the center of different size cubolds of interspersed moderation' depending on the spacing between fuel assemblies. The same interspersed moderation was used inside the fuel assembly as that surrounding it. By using reflection on all sides, an infinite array of fuel assemblies was computed. When a finite array of assemblies was calculated, the assembly was homogenized using the XSDRNPM code in the CSAS2 module and placed into a finite array of fuel assemblies with vacuum boundary conditions on all faces. The interspersed moderation was varied from 0% to 100% dense water. Both 4.1 wt% and 5.1 wt% Um enrichments were calculated.
The fuel was assumed to be 100% theoretical density UO, in all cases to provided additional conservatism to the calculations.
Page: 15-21 December 30,1996 Revision: 5
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS 15.2.6.2 Fual A=cambiv Storage calculational Ranulta CASE #1 The 17 x 17 MKRW type fuel assembly was run with 5.1 wt% U;. and 100.0% theoretical density fuel. The assenibly was modeled in detail showing Individual pins. The assembly was reflected in both the X and Y directions to obtain an infinite array of assemblies with spacing of 21" center to-center (CTC) in one direction and 38" center-to-center in the other direction. Two interspersed moderation cases were run. With 0% dense water (normal conditions) the infinite array km was 0.383 i.003. With 5% dense water for the interspersed moderation in and around the assemblies, the k,,,#, was 1.086 i.005. The 5% dense interspersed moderation gives a k,,,#,
that is unacceptable for an infinite array spaced at 21" X 38" CTC. For 4.1 wt% enrichment and 5%
l dense interspersed moderation, the k,,,#, was 1.054 i.005. The actual assembly storage array j
is not an infinite array but is actually a 10 X 20 array of storage positions spaced at 21" CTC in l
the 10 assemblies direction and 38" CTC in the direction with 20 assemblies. The total storage array has 200 fuel assembly positions built into the array.
CASE #2 Both 5.1 wt% and 4.1 wt% fuel assemblies were modeled with 5% dense interspersed moderation
)
into the actual 10 X 20 array with 21" X 38" spacing. The 4.1 wt% assemblies k,,,#, was 0.937 i.006. The 5.1 wt% assemblies had a k,,,#, of 0.970 i.005. The 5% interspersed moderation gives the highest k,,,#, for the 21" X -
38" CTC spacing. Both the 4.1 and 5.1 wt%
enrichment give k,,,#,,, that exceed the allowable 0.95 when the 2 sigma and 0.02 bias is added. It is seen that the 4.1 wt% enrichment is very close to being acceptable. K.,,#, values improve rapidly both above and below the optimum 5%
dense interspersed moderation condition. It is not Page: 15-22 December 30,1996 Revision: 5 i
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS 1
considered a credible situation to have the entire 10 X 20 array of fuel assemblies at the optimum 5% dense moderation with the way the array is
)
made and with the given conditions on the use of hoses and water lines in the area. A more l
comprehensive discussion of the incredibility of having the entire array at the 5% dense optimum moderation will follow later.
1 i
CASE #3 A 10 X 10 array of fuel assemblies (half the actual storage array) was calculated at 4.1 wt%
i enrichment with various degrees of interspersed j
moderation. The results is in Table 4. The 5.1 wt% enrichment was calculated at the limiting j
(5% moderation) condition.
1 4
I Table 4.
K-effective for a 10 X 10 Array of Fuel Assemblies i
at Various Interspersed Moderation.
Percent 1
Interspersed Enrichment K-effective i 2 sigma j
Moderation Wt% U-235 i
l 4
4.1 0.891 i.005 j
5 4.1 0.910 i.005 1
6 4.1 0.904 i.005
}
100 4.1 0.894 i.006 5
5.1 0.943 i.006 1
1 l
It is seen that the 4.1 wt% enrichment is below the allowable 0.95 k-eff for all degrees of j
moderation but the 5.1 wt% is not acceptable at the 5% moderation even for the 10 X 10 array slZO.
1 l
CASE #4 An infinite array of 5.1 wt% assemblies was calculated for a spacing of 42" X 38". This assumes that every other location is used in the i
i j
Page: 15-23 December 30,1996 Revision: 5 i
s
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS l
21" spacing direction. This give a total storage array of 5 X 20. The infinite array km for 5.1 l
wt% was 0.897 5. 005. This shows that all enrichments below 5.1 wt% and at all degrees of l
interspersed moderation is below the allowable 0.95 km accident condition. The 5.1 wt%
l enrichment was calculated for various degrees of l
interspersed moderation. The results of these calculation are shown in Table 5.
CASE #5 A calculation was made for 5.1 wt% enrichment fuel with the 42" X 38" CTC to determine the 1
l effects on k,#,if assemblies were to accidentally fall together while in the normal dry condition. A 4 X 4 array of fuel assemblies was 1
modeled so that either 2 or 4 assemblies could be
{
placed next to each other. The use of reflection boundary conditions in the X and Y directions made this an infinite array. The base k, for the 4 j
X 4 array with all assemblies spaced normally at j
42" X 38" was 0.269 &_.002. The k,#, for having 2 assemblies fall together was 0.267 &_
.002 which is a slight reduction in k,
. The j
k,#, for the case of 4 assemblies falling together l
was 0.275 5. 003 which is well below the acceptable 0.95 k-effective for accident conditions.
Radioloalcal Safetv
\\
The Fuel Assembly Storage Area is an uncontrolled area that processes fully clad SNM in the Fuel Assembly configuration. Those radiological controls used for clean area are appropriate for the assembly storage area also. There have been no contamination or internal personnel exposure problems in this area. TLD's are used to measure the external exposure levels.
15.2.6.3 Credibility of Achievino Ontimum Moderation Achieving optimum moderation levels (0.05 gm/cc) within any portion of sufficient size within the 21 x 38 Page: 15-24 December 30,1996 Revision: 5
FRAMA TOME COGEMA FUELS - L YNCHBURD MANUFACTURIN3 FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS
}
1 l
inch assembW storage array (>._100 assemblies) to result
{
in K-eff > 1.0 is not considered credible under existing j
and anticipated conditions.
t
[
f The following points were considered in this ovaluation:
l 1
ll No steam sources are located near the storage array.
No sprinklers are located near the storage array.
i l
A finite 100 fuel assembly array at the 21 x 38 inch spacing has been demonstrated to be safe for i;
enrichment below 4.1 wt%. The probability that a l
l solid stream of water (for example, a fire hose) di-1 rected into the array would affect even 100 l
assemblies is so remote as to be negligible. As a i
further conservatism, the use of hoses to fight fires within the storage array will be prohibited unless prior authorization has been obtained from j
a knowledgeable management representative of j
the plant emergency response organization.
i t
i If thn use of a fire hose is considered necessary, no more than one line will be applied to the i
storage array at any given time.
I Restrictions on the use of hose lines shall be l
prominently posted at the boundaries of the array i
such that they are visible from all fire fighting l
approach routes.
i At the present time, three (3) water lines penetrate j
the assembly storage area. These are low i
pressure lines with a maxirnum 2 inch diameter.
l The lines are located along the external plant walls l
approximately 14 feet above the floor (i.e., at the j
same height at top of the fuel assemblies in i
storage configuration). In the event a rupture
]
should develop, the resulting water stream would i
l 1
Page: 15-25 December 30,1996 Revision: 5 a
l 1
i
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS i
l interact with a very limited number of assemblies j
due to the low volume and pressure of the water.
Flammable material accumulation in the area is i
minimal due to the nature of the operation. The lack of flammables, coupled with high quality 4
housekeeping standards result in a minimal fire hazard.
i 15.2.6.4 Nuclear Interaction i
Nuclear interaction between fuel assembly processing and assembly storage is limited to less than the l
Interaction within the storage array alone due to j
geometric relationship. Since a planar array of assemblies on 21 inch centers is subcritical for credible interspersed moderation conditions, the planar array in I
the assembly processing area (where four foot separation is maintained) is likewise subcritical. Two or more planar j
4 arrays have been shown to be subcritical in combination j
i when center-to-center distance is not less than 38 j
inches, and with major faces opposed, the situation j
resulting from considerations of the storage array.
Nuclear interaction between assembly storage and assembly shipping container loading is limited to the interaction associated with an array of damaged loaded shipping containers. The evaluation of damaged shipping containers shows that the containers are nuclearly safe with minimum possible fuel assembly edge-to-edge spacings of 7,8 and 18 inches. The fuel assembly a
storage rack is also shown to be safe based on 21" x 38" center-to-center spacing for 4.1 wt% and below and j
42" X 38" spacing for enrichments above 4.1 wt% but below 5.1 wt% as previously stated in the criticality evaluations. The minimum required edge-to-edge i
spacing of 18" between assemblies in adjacent shipping containers (see Chapter 4, Section 4.2.4.7 reference 1) corresponds to a 38" center-to-center spacing between the loaded adjacent Model B containers. Thus,38 inches edge-to-edge separation between the fuel assembly 4
Page: 15-26 December 30,1996 Revision: 5
' FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY 1
USNRC LICENSE SNM-1168, DOCKET 70-1201
^ -
PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS.
1 i
storage array and the fuel assembly loading area is conservative and will provide acceptable nuclear interaction control.
4 Rotation of one planar array 90 degrees from the vertical will present its edge to the face of the opposing array, while maintaining the specified separative distances.
j This significantly reduces interaction probabilities by presenting less " effective" surface area for interaction 4
i which, in effect, extends the effective separation distance. This demonstration accurately reflects j
conditions at the junction of the fuel assembly l
processing and storage area, and also at the junction of the assembly storage and shipping container loading area.
Under normal operating conditions, the activities will be l
conducted in the absence of any significant moderation.
Fuel of 5.1 wt.% enrichment under such conditions and less cannot be made critical.
15.2.7 Fresh Fuel Shloolna Container Criticalitv Safety i
Nuclear safety analysis for the shipping container determined j
that an infinite array of damaged containers (with seven to eighteen inches separation distance'between assemblies) was f.
subcritical through all degrees of water moderation. Likewise, j
undamaged containers with a minimurn of twenty-four inches separative distance between assemblies is subcritical. Storage of containers under the constraints imposed in Chapter 4 assures that no more reactive condition than that considered can be attained. Consequently, accidental criticality is effectively precluded by assuring that conditions are as evaluated and remain so. Subpart D of 10 CFR 71 has been 4
designed to provide assurances regarding the condition of the container and its contents. The application of seals in addition i
to protection from unauthorized handling assures the continuance of safe conditions. The absence of nuclear 1
interaction with other fissile materials to cause accidental criticality is assured by providing nuclear isolation.
Page: 15-27 December 30,1996 Revision: 5 I
- - ~. -....
l FRAMATOME COGEMA FUELS - LYNCHBURG AGANUFACit/ RING FACILITY l'
USNRC LICENSE SNM-1968, DOCKET 70-1201 l
PARTM - CHAPTER 15 - PROCESS DESCRIP110N AND SAFETY ANAL YSIS i
4 Loaded containers will be stored in the north and of the main s
j LMF plant or on a pad adjacent to the north end inside the security fence.
l i
in consideration of the foregoing, FCF requests exemption from t
the criticality monitoring requirement as provided for in 10 CFR 3
j 70.24(d).
Nuclamr Interaction i
i l
As previously demonstrated, 38 inches edge-to-edge separation J
j between the fuel assembly storage array and the fuel assembly i
loading area is conservative and will provide accepte51e nuclear
}
interaction control. No other form of special nuclear material is i
located closer to loaded shipping containers than loaded fuel assembly racks. This applies to' any currently authorized j
l enrichment at moderation levels equal to, or greater than 10%
l when reactivity of the units does not exceed that specified in
)
i Table IV, TID-7016 (Rev.1). At less than 10% moderation, interaction between the shipping containers and other arrays j
becomes insignificant due to rapidly decreasing values of K-eff.
i j
calculational Model Descrintion The criticality safety analysis for fresh fuel shipping container j
was performed using the KENO-IV Monte Carlo code. Infinitely dilute cross sections were obtained from the 123 XSDRN master library cross section set. The NITAWL code was used to generate a 123 group KENO-IV working library that included resonance self shielding of the isotope Uranium-238, the only resonance absorber present in the fresh fuel. The fuel lattice rod-to-rod self shielding for the U-238 resonances was accounted for with a Dancoff factor by NITAWL. The Dancoff factor was generated by the B&W NULIF code, a neutron spectrum generator and spectrum weighted few group constant calculator.
Three dimensional heterogeneous geometry models were used for all KENO-IV criticality analyses. Individual components of the fuel assembly lattice were modeled along with the poison plates adjacent to each fuel assembly, surrounding moderator Page: 15-28 December 30,1996 Revision: 5
FRAMATOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201
\\
i PARTH - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS i
1 regions and container shell. Moderator was substituted for fuel
]
assembly grids, and fittings and structural steel components as l
a conservative geometry modeling simplification.
The accident condition package array was modeled as an infinite (symmetry boundary conditions) XY crushed container j
array of one active fuel column length (144 inches) with eight (8) inches of water on each end. For the most reactive j
moderation condition,100% waiter moderator and water j
reflector, KENO calculations were made for both an eight and twelve inch water reflector at each end of the active fuel stack.
There was no significant difference in the keff due to the end
{
reflector thickness.
The normal condition package array was modeled as an infinite l
(symmetry boundary conditions) XYZ container array with a 24 inch (minimum) separation between fuel assemblies of different containers. For this analysis, the container was replaced by l
moderator.
j Fuel Assembiv - The specification for the assembly is the B&W low enriched PWR fuel assembly product line with 5.1 wt.%
i U-235 and 96.3% T.D. pellet density. All control rods and i
instrument tubes are removed; the control rod guide tubes and i
instrument guide tube are present.
i Shinnino Container - The container holds two assemblies and is crushed to simulate an accident condition. The relative location of items with the container is unchariged from the normal case.
l The container is internally poisoned by two stainless steel plates containing 1.5 wt.% boron. See Figure 15.6.
l Arrav - The containers form an infinite array in two dimensions, l
one container long. Eight inches of water reflection is assumed l
on the container ends.
li Moderation - The amount of moderation is not controlled but is assumed to be a uniform density of light water.
l Radioloalcal Safety 1
i f
Page: 15-29 December 30,1996 Revision: 5 i
4
.- - -_ = -.
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACill1Y USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS The radiological concerns of the shipping container are minimal.
The container is used in uncontrolled areas with fully clad SNM in the Fuel Assembly configuration. No real potential for airborne contamination or internal personnel exposure. Clean area controls are appropriate for this area. External exposure levels are monitored via TLD badges.
I 15.2.8 UF, Storage j
Stornoe Conditions i
Natural UF., as received for storage only, will be in solid form and stored in cylinders that are approved for shipment. The cylinders will be stored outside, in an area designated for UF.
storage within the LMF security fence. The cylinders will be i
stored on timbers designed to cradle and give support, in an j
. area prepared and maintained under weed control, and located 1
a minimum of 100 ft. from the paved driveway which circles j
the LMF. The cylinders will be equipped with valve protectors, and will be routinely monitored by Health-Safety personnel to detect leakage and to assure the continued adequacy of j
storage conditions. Outside storage of UF, cylinders has been i
previously determined to have no adverse environmental impact l
(Reference 6).
UF., Criticality Safety - Special Nuclear Material i
Not applicable as only natural UF, not enriched is authorized to 1
be accepted at LMF.
i 15.2.9 Miscellaneous criticality controls 1
Laboratorv And Develonment Ooerations j
j Figure 21 of TID-7028 indicates that a 5.1% enrichment 700 grams of U-235 is approximately 45 percent of the minimum critical mass under optimum conditions.
4 Although it has been established in Part 7.10 that 700 grams of U,3s represent a safe mass for 5.1% enriched U,3s under J
Page: 15-30 December 30,1996 Revision: 5
-.. - =
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS optimum conditions,350 grams is selected for laboratory and development purposes to assure absolute safety with maximum flexibility for the nature of the work.
1 j
Contaminated Solid Wasta j
The quantity of SNM adhering to solid waste as contamination is sufficiently small to completely preclude the collection of even 350 grams of Uru in any waste container.
15.2.10 References i
3 1.
R. D. Carter, Criticality Handbook, Volume 1, ARH-600, June 28,1968.
j 2.
J. H. Chalmers, Handbook of Criticalltv Data. United i
Kingdom Atomic Energy Authority,1965, (AHSB).
)
3.
H. K. Clark, " Critical and Safe Masses and Dimensions of 1
Lattices of U and UO Rods in Water", DP-1014, 2
February 1966.
k j
4.
H. C. Paxton, " Criticality Control in Operation with Fissile l
Material", LA-3366, January 15,1966.
i 5.
Stevenson-Odegaarten, " Studies of Surface Density Spacing Criteria using KENO Calculations", ANS-TRANS l
12, 890, November,1969.
i 6.
BAW-1412, Annex 1; UF, to UO Conversion Facility.
2 i
1 i
Page: 15-31 December 30,1996 Revision: 5 i
f a
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURIN3 FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS 1
l FIGURE 15.1 U02 RECEIPT NON-GNM RECEIPT (i. e.
CLADDING I
(- C BAR STOCK etc)
WEIGHING / SAMPLING O
I I
N I
TA CLADDING PREP NON-StM PARTS PELLET RR (UT, DIMENS.3 FABRICATION STORAGE VAULT OE END CADS, GRIDS, i
LA SPACERS,END I
I L
1st END CAP FITTINGS, ETC.
ROD LOADING AREA E
(- 0
(-
I
(
(
l DRYING RETORTS l
He PRESSURIZATION LASER WELD i
ROD SCANNER I
CLEAN!NG ROOM l
ALPHA COUNT I
FINAL INSPECTION l
He LEAK TEST l
CHANNEL STORAGE I
F/A ASSEMBLY
(
F/A STORAGE l
F/A SHIPPING Page: 15-32 December 30,1996 Revision: 5
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURIND FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARYM - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS FIGURE 15.2 8" CONCRETE CEILING n
111" a
FUEL BOX 8
s, ss
/.
N 0.25" Borel O.25" Boral M4 C
16" C
H H
C C
ss 0
0 N
O.25" Borel O.25" Borel M-9 I
C C
R R
E E
T T
n ss E
E 0.25" Borel O.25" Borel U
U A
A L
L
<i 0.25" Boral O.25" Borel i
18" I'
36" 18" i
l 0.25" Boral lo" O.25" Boral h
12 " CONCRETE FLOOR (thickness assumed)
I f
PDQ - 07 Grid l
1 l
l i
Page: 15-33 December 30,1996 Revision: 5
{
i
FRAMA TOME COGEMA FUELS - LYNCHBUR3 MANUFACTURIN3 FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANAL YSIS FIGURE 15.3 K-EFFECTIVE VERSUS VOLUME FRACTION FUEL 4" THICK INFINITE SLAB OF CLADDED RODS 0.94 C
l 0.88 O.86 0.84
\\
La 2
0.82 g
g 1
0.78 0.76 x
0.74 0.72 O.7
'A 0.68
+u 0.66 0.2 0.3 0.4 0.5 0.6 0.7 VOLUME FRACTION OF FUEL O
MKB O.370" PELLETS
+
MKBW O.3195" PELLET Page: 15-34 December 30,1996 Revision: 5
FRAMA TOME COGEMA FUELS - LYNCHBUR3 MANUFACTURIN3 FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 15 - PROCESS DESCRIPHON AND SAFETY ANAL YSIS FIGURE 15.4 K-EFFECTIVE VS INTERSPERSED MODERATION CNFP VAULT KENO DATA O.95 0.9 Q
a 0.85 0.8
${
O.75 E
O.7 xer 0.65 0.6 0.55 0.5 0
20 40 60 80 100 PERCENT INTERSPERSED MODERATION Page: 15-35 December 30,1996 Revision: 5 1
FRAMA TOME COGEMA FUELS - L YNCH:UR3 MANUFACTURIN3 FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS RGURE 15.5 i
K'-EFFECTIVE VERSUS PERCENT IM FUEL ASSEMBLY FABRICATION 1
0 e
0.9 i
/
0,8 f
0.6 g
E b
05
/ /
0.4 Y
0.3 0.2 0.1 0
20 40 60 80 100 PERCENT INTERSPERSEO MODERATION O
48" ETE
+
38" CTC Page: 15-36 December 30,1996 Revision: 5 1
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 j
PART11 - CHAPTER 15 - PROCESS DESCRIPTION AND SAFETY ANALYSIS 1
i FIGURE 15.6 i
O.180 thick Carbon Steel 1
j l\\
/
.s...
/
f J
2 - 11/16" gap g
w/ Boran S.S.
J n-m m
II
\\
I/
N 8-I I
i Page: 15-37 December 30,1996 Revision: 5
4 FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 16 - ACCIDENTANALYSES 16.1 GENERAL l
An gnalysis of the LMF facility and processes was conducted to identify.
l and quantify the potential environmental impact of postulated conceivable accidents. Since LMF operations involve UO,in pellet form only, the consequences of an accident at LMF that has an environmental impact is not very probable except in the case of a criticality. In the event of a criticality, the major environmental impact would result from the release of fission products, in order for the fission prt, ducts to be dispersed to affect the general public, we would have to assume that our ventilation system was operating but the HEPA filters were degradated.
The pellet loading room (PLR) and vault is the only area where SNM is handled that requires a HEPA ventilation system. Therefore, it is was 4
chosen as the area that would produce the greatest impact on the environment if a criticality were to occur. Although, the fuel storage rack's SNM inventory is the largest in the plant, if a criticality were to j
occur the fission products would not be discharged off-site since there is no ventilation system associated with the area, l
This analysis followed the assumptions and conditions proposed by the NRC Regulatory Guide 3.34 and 3.35. The effects of the postulated accident should be considered as conservative with regard to the present scope and operation of the LMF. As a result, the report is useful and l
3 serves the purpose of conservatively evaluating the effects of a criticality accident at the LMF.
This chapter summarizes the results of the analysis performed in accordance with the NRC Reg. Guide 3.34 and 3.35. The major process or equipment differences will be noted in the discussion.
16.1.1 Maximum Credible Accident Imoact An analysis of the LMF facility and processes was conducted to l
identify the potential environmental impact of postulated
]l conceivable accidents. In summary, the maximum impact from any of the postulated accidents on the population outside the site boundary was determined to result from a criticality accident.
q lt was calculated that this highly improbable accident could result in a dose to an individual at the site boundary (250 meters) to be 3.5 rem whole body based on a postulated % hour exposure. At Page: 16-1 December 30,1996 Revision: 3
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 16 - ACCIDENTANALYSES 1
l the nearest residence, % mile away (800 meters), the individual dose for an X hour exposure was calculated to be 8./ rem whole 4
body. These values may be compared to the accident doses for establishing exclusion areas, given in 10 CFR 100.11, of 25 rem whole body. Consequently, the postulated maximum credible accident would produce doses significantly less than the design doses specified in 10 CFR 100 for reactors.
l The following are assumptions or conditions of the accident as j
used in the analysis:
i The results should be considered as upper limits of the consequences of the postulated accident.
i' The consequences of accidents in Pellet Loading Room area are many orders of magnitude below the impact of accidents in other types of nuclear facilities.
l 4
The accidents selected for analysis in the report are conceivable but very unlikely due to the engineered and administrative controls employed at the LMF.
l 16.1.2 Classes of Accidents i
There are four classes of accidents designated as being
{
representative of those that could occur at the LMF. These are j
outlined in the following discussion. Where appropriate, the j
following topics are discussed for each accident:
i j
- Possible causes and their probability of occurrence
- Engineered safety features provided
- Nature and a.rount of radioactive effluent released to the environment
- Effects of dilution in the environment Page: 16-2 December 30,1996 Revision: 3
4 i
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 a
PARTM - CHAPTER 16 - ACCIDENTANALYSES I
l 1
- Transport and effects to the nearest population and the maximum individual
)
4 i
e Time (scale) sequence of the postulated accident and of l
possible adverse effects 1
ACCIDENT CLASSES Class 1 -
a j-Trivial incidents - No Releases to Environment i
l
- Ventilation System Failures i
e Loss of Electrical Power Class 2 -
Small Environmental Releases
- HEPA Filter failure I
j Class 3 -
i f
Maximum Credible Accidents
- Major Fire l
- Major Explosion j
- Tornado
- Criticality i
e Transportation Accidents Class 4 -
j Very Remote Hvoothetical Accidents (Reactor Class 9) 4 1
4 Page: 16-3 December 30,1996 Revision: 3 i
l
1y, FRAMA TOME COGEMA FUELS - L YNCHBURG ntANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201
~
PARTM - CHAPTER 16 - ACCIDENTANALYSES i.
j i
16.1.2.1 CLASS 1 - TRIVIAL INCIDENTS l
l A number of smaller scale incidents are reviewed to assure that they would not cause significant j
environmental irnpact. A trivial incident generally has a relatively high probability of occurring, but j
results in no significant release of radioactive material to the environment. Such incidents j
include electrical power outage, loss of water supply, ventilation system mechanical failure, and
)
minor spills within the plant. No equipment was identified from a plant inspection and a review of i
facility drawing for which loss of power or water l
would cause accidental environmental impacts
{
different from those analyzed in other sections.
Mechanical failure of ventilation equipment would still leave HEPA filtration to prevent release to the environment. Minor uranium spills in the plant would be within the containment offered by the HEPA filters and no release to the environment l
would be expected.
1 i
16.1.2.2 CLASS 2 - SMAlI ENVIRONMENTAL REL FASE The greatest potential for a small environmental i
release would be if the HEPA filters in the i
ventilation system failed. Since the filter system i
has several redundancies inherent in its design, l
this could only be achieved through a major l
i accident, in this event, the system would be shut j
down. Hence, there is no potential for a small environmental release at the LMF.
l 1
i i
i Page: 16-4 December 30,1996 Revision: 3 i
.h i
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1968, DOCKET 70-1201 PARTH - CHAPTER 16 - ACCIDENTANALYSES 16.1.2.3 CLASS 3 - MAXIMUM CREDIBLE ACCIDENTS 16.1.2.3.1 Malor Fire if there was a major fire at LMF, the greatest potential for an offsite release would be in the pellet loading room for the likelihood that the ventilation system would be degraded. However, the dose to the site boundary and the nearest residence would be insignificant since UO powder is no longer present and 2
only UO, pellets are used in the facility. Also, in the event of a major fire, fire fighting techniques mandate that electrical power is shut-off which would disable the ventilation system from discharging offsite.
16.1.2.3.2 Malor Exolosion There are no significant potential for a major explosion at the LMF. No explosive materials are present or stored in any significant amount to cause a major explosion.
16.1.2.3.3 Earthouake i
The LMF is located in an area classified as Zone 2 on the Seismic Risk Map of the United States and corresponds to an intensity of Vil on the Modified Mercalli Scale. This intensity has an acceleration range of 0.06 to 0.14 g and implies variable Page: 16-5 December 30,1996 Revision: 3
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 16 - ACCIDENTANALYSES damage to buildings, depending on coristruction.
It is assumed there would be no loss of integrity of the metal frame, sheet-metal-covered LMF facility, due to the metal's ductility. Shifting and toppling (especially of top-heavy items) could occur.
No release of uranium dioxide to tha atmosphere could result from any proposed sequence of events. The uranium dioxide is present in the form of pollets only.
16.1.2.3.4 Tomndo The LMF is not designed to withstand the direct impact of a tornado.
However, the LMF is located in a i
relatively low probability area for tornadoes in the United States. The probability _of a tornado actually striking the site in any given year is estimated to be 3.0 x 10~', with a recurrence interval of 3333 years.
In the event of a tornado, there will be no release of any significant amount of uranium dioxide to the atmosphere. The UO, pellets are not likely to be airborne and can not be Inhaled.
l Page: 16-6 December 30,1996 Revision: 3 l
1 1
l, FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY l
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 16 - ACCIDENTANALYSES l
16.1.2.3.5 Criticalitv The location that would have the largest environmental impact from a j
criticality would be in the pellet loading room (PLR) and vault area.
l All other areas of the plant is not i
ventilated to the environment so that j
the radioactivity would remain on-site.
i f
Criticality evaluations performed in
]
Chapter four illustrate that when'the l
pellets are maintained in a safe configuration (four inch slab height and twenty-five inch width) that even under optimum moderation conditions a criticality cannot be achieved.
Therefore, in order to sustain a criticality, there must be an upset to the integrity of the vault shelves and moderation must be introduced to allow the proper mix of UO and 2
water to propagate a criticality accident. The circumstances of such an accident occurring cannot be conceived. However, the analysis of a such an unlikely event is described next.
Due to the low fissile content of UO 2 pellets handled at the LMF, criticality with unmoderated (dry) accidental i
accumulations is not highly credible.
For example, the critical mass of 5 gm/cc powder with a hydrogen-to-
)
uranium ratio of 0.6 or less (2 wt.%
water) would be greater than 100,000 kg. Therefore, credible Page: 16-7 December 30,1996 Revision: 3 j
FRAMA TOME COGEMA FUELS - LYNCHRURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PART11 - CHAPTER 16 - ACCIDENTANALYSES i
accident conditions leading to a criticality would include the provision of neutron moderating materials.
Design of the LMF is such that significant neutron moderating material is excluded from plant working areas. For instance, water service lines are placed on the outer walls of the building and are covered.
Also, many of the processing steps in the LMF would be safe even if fully moderated and reflected, providing i
extra conservatism when procedures j
and design exclude availability of moderating materials.
i The minimum volume of fuel / water mixture which will achieve criticality at 4.0 wt.% enriched UO, is estimated to be approximately 26 i
liters. Accident conditions must, l
therefore, provide not only a sufficient amount of fissile material (115 kg U at 4.0 wt.%) and water l
(26 liters), but also a contained l
volume of approximately 26 liters to hold such a mixture. The most i
credible location for such an event is the PLR and vault area.
3 The accident conditions chosen for i
analyzing a criticality event should be those which would yield the maximum credible environmental impact. Most effects of a criticality are localized to the plant and its occupants, except for any formation and release of fission products.
i i
Page: 16-8 December 30,1996 Revision: 3 1
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 16 - ACCIDENTANALYSES Therefore, the maximum impact would occur from accidents in the PLR vault area. The assumptions and conditions of this proposed accident in the PLR area are adopted from the Regulatory Guide 3.34 and 3.35.
i The probability of this accident to take place should be considered very unlikely and a worst case scenario.
The impact of a criticality event based on such a scenario is presented in this section.
This impact was calculated using a criticality event as proposed in the Regulatory Guide 3.34 and 3.35. The number of fissions for the first pulse are 1x10 fissions in 0.5 seconds followed successively at 10-minute intervals by 47 bursts of 1.9x10" fissions for a total of 1x10 fissions in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This number of fissions is as high or considerably higher than has been experienced in previous criticality accidents. The total yield conservatively rapresents the scenario condition that an outside source would remoderate the system after i
heating has damped the first criticality by expansion and expulsion of some moderator, it should be possible to shut off water lines to the plant in i
less than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> represented by the postulated event. Also, LMF monitoring instruments should detect a criticality and alarms should signal an immediate evacuation of the plant,
)
l Page: 16-9 December 30,1996 Revision: 3
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTll - CHAPTER 16 - ACCIDENTANAL YSES thus reducing the % hour exposure for an Individual at the plant boundary. Environmental impact from the postulated criticality accident should, therefore, be no greater than
)
presented in the table below, and l
would most likely be considerably less.
4 Doses from Postulated Criticalitv Accident inhalation Doses
(
Individual 4
At Plant Boundary 250 m - % hr.
exposure
- 1. Total Body - 3.24 rem At nearest Residence % mile (800 m)
- % hr. exposure
- 1. Total Body - 8.69 rem External Doses i
jndividual At Plant Boundary 250 m = % hr.
~
exposure
- 1. Total Body - 0.2985 rem At Nearest Residence % mile - % hr.
exposure Page: 16-10 December 30,1996 Revision: 3
.. _. ~.-.-...
_. -. = ~
FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACITJRING FACILITY I
USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 16 - ACCIDENTANALYSES
- 1. Total Body - 0.0025 rem 16.1.2.3.6 Transnortation Accidents j
Radioactive materials are transported both to and from the LMF. Materials l
shipped to the plant consist primarily of uranium dioxide in the form of solid sintered fuel pellets. Shipments from the facility will consist primarily of
)
unirradiated finished fuel assemblies composed of sintered UO, pellets encased in zirconium alloy tubes.
Present plans consider all shipments l
will be made by truck.
All radioactive material shipments are subject to the stringent regulations and requirements of Nuclear Regulatory Commission and the Department of Transportation. These i
regulations specify that the shipping packages must be designed to withstand certain specified normal conditions of transport and i
hypothetical accident conditions I
without loss of contents, significant loss of shielding, and (in the case of 1
fissile materials) without criticality.
l I
l Based on regulatory standards.and l
requirements for package design and quality assurance and the results of tests and past experience, these packages are designed to withstand l
all but the very severe, highly unlikely accidents. The probability of a l
package being breached is so low
)
Page: 16-11 December 30,1996 Revision: 3 l
4 FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTM - CHAPTER 16 - ACCIDENTANALYSES that a detailed evaluation was not considered necessary. In addition, the consequences associated with a release of UO pellets are quite small 2
and the probability of occurrence is also small; therefore, the risk of impact to the environment is very small.
16.1.2.4 Very Remote Hvoothetical Accidents Accidents in this category are of such a low probability of occurrence, either because of design safeguards or the low frequency of casual events, that a detailed environmental impact assessment was not attempted.
However, they were given some consideration to make certain that the detailed assessment did not overlook any other significant potential environmental impact. In all such cases the consequences of these postulated accidents would not exceed those of the previously considered criticality.
Maximum Flood The effects on the LMF a large flood would be minimal.
The 500 year flood would have a stage of 497 feet, while the floor of the LMF is 547 feet. The Standard l
Project Flood proposed by the Corps of Engineers would reach to 502 feet, still 45 feet below the LMF floor.
Consequently, the possible effects of flooding at the LMF are considered to be zero.
Destructive Earthauake An earthquake might be considered which would cause the release of the total UO, pellets inventory rather than a small fraction. If this were the case the consequences l
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 16 - ACCIDENT ANAL YSES b
will have no impact on the population close to the plant since the pellets can not be inhaled.
Other Accidents 1
Other accidents which might be postulated range from an airplane impact into the facility, with a probable i
frequency of occurrence per plant year of 10-5, to a meteorite strike with a probability of 10 ' per year.
4 These accidents should not cause any significant
)
environmental or radiological consequences.
l 4
4 l-i i
Page: 16-13 December 30,1996 Revision: 3
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