ML20236U712

Application for Amend to License SNM-1107,revising Page IV & Chapters 1.0,2.0,3.0,5.0 & 8.0

ML20236U712
Person / Time
Site:	Westinghouse
Issue date:	07/23/1998
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	NRC
Shared Package
ML20236U710	List: ML20236U709 ML20236U712
References
NUDOCS 9807300392
Download: ML20236U712 (77)

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{{#Wiki_filter:_ - _ _. _ _ 0: TABLE OF CONTENTS NUMBER AND TITLE PAGE TABLE OF CONTENTS............................................................................ REVISION RECORD................................................................................ CHAPTER 1.0 GENERAL INFORM ATION.................................................. 1.0 1.1 - FACILITY AND PROCESS DESCRIPTION.............................. 1.0 1.2 IN STITUTIONAL INFORM ATION......................................... 1.4 1.3 SITE D ESCRIIrrION............................................................ 1.5 1.4 TERMS AND DEFINITIONS................................................. 1.8 CHAPTER 2.0 MANAGEMENT ORGANIZATION........................................ 2.0 2.1 ORGANIZATIONAL RESPONSIBILITIES AND AUTHORITIES....................................................................- 2.0 2.2 S AFETY COMMITTEES...................................................... 2. 8 CHAPTER 3.0 CONDUCT OF OPERATIONS............................................... 3.0 3.1 CONFIGURATION MANAGEMENT...................................... 3.0 l 3.2 MAI NTENANCE............................................................... 3.2 3.3 Q UALITY ASSURANCE...................................................... 3.4 3.4 PROCEDURES, TRAINING AND QUALIFICATION................. 3.8 3.5 H UMAN FACTORS............................................................ 3.14 3.6 AUDITS AND f. '.F-ASSESSMENTS.................................... 3.15 3.7 - INCIDENT INVLaTIGATIONS............................................. 3.17 3.8 RECORDKEEPING AND REPORTING.................................. 3.21 CHAPTER 4.0 INTEGRATED S AFETY ASSESSMENT.................................... 4.0 CHAPTER 5.0 - RADIATION SAFETY........................................................ 5.0 5.1 ALARA (As Low As Reasonably Achievable) POLICY................. 5.0 5.2 RADIATION WORK PERMITS (RWP).................................... 5.1 5.3 VENTILATION SYSTEMS.................................................... 5.2 5.4. AIR S AM PLING................................................................. 5.4 5.5 CONTAMIN ATION CONTROL............................................. 5.5 i 5.6 EXTERNAL EXPOS URE...................................................... 5.8 i l 5.7 INTERNAL EXPOSURE....................................................... 5. 8 i

5.8 RESPIRATORY PROTECTION............................................ 5.11 l.

5.9 INSTRUMENTATION......................................................... 5.12 l' 5.10 SUMMING INTERNAL AND EXTERNAL EXPOSURES........... 5.12 Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. _ i ) 1 License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 9007300392 900723 PDR ADOCK 07001151 i C PDR 1 m. J

e. TABLE OF CONTENTS (Cont'd) NUMBER AND TITLE PAGE CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY........................................ 6.0 6.1 PROGRAM ADMINISTRATION............................................ 6.0 6.2 CONTROL METHODOLOGY AND PRINCIPLES.................... 6.8 6.3 ALARM SYSTEM.............................................................. 6.1 1 6.4 CONTROL DOCUMENTS................................................... 6.21 6.5 ALARM S YSTEMS............................................................ 6.22 CHAPTER 7.0. CHEMICAL S AFETY.......................................................... 7.0 7.1 CHEMICAL SAFETY PROGRAM......................................... 7.0 7.2 CHEMICAL SAFETY HAZARD EVALUATIONS...................... 7.0 7.3 CHEMICAL SAFETY PROGRAM STRUCTURE....................... 7.1 7.4 ADDITIONAL CHEMICAL SAFETY COMMITMENTS.............. 7.2 CHAPTER 8.0 FIRE SAFETY.................................................................... 8.0 8.1 STRUCTURE OF THE FIRE SAFETY PROGRAM..................... 8.0 8.2 FIRE SUPPRESSION SERVICES.......................................... 8.10 CHAPTER 9.0 EMERGENCY MANAGEMENT PROGRAM........................... 9.0 9.1 EMERG ENC Y PLAN........................................................... 9.0 9.2 EMERGENCY EQUIPMENT................................................. 9.0 CHAPTER 10.0 ENVIRONMENTAL PROTECTION...................................... 10.0 10.1 EFFLUENT AIR TREATMENT........................................... 10.0 10.2 LIQUID WASTE TREATMENT FAC'LITIES........................... 10.0 10.3 SOLID WASTE DISPOS AL FACILITIES................................. 10.1 -10.4 PROGRAM DOCUMENTATION........................................... 10.1 10.5 EVALUATIONS................................................................ 10. 2 10.6 OFF-SITE DOS E................................................................ 10.2 CHAPTER 11.0 DECOMMISSIONING........................................................... 1 1.0 11.1 CONCEITI'UAL DECOMMISSIONING PLAN.......................... 11.0 11.2 DECOMMISSIONING FUNDING PLAN AND FINANCIAL ' ASSURANCE................................................................... 1 1.1 CHAPTER 12.0 AUTHORIZATIONS AND EXEMIrI' IONS............................... 12.0 12.1 AUTH ORIZATIONS........................................................... 12.0 l 12.2 EXEMPTI ONS................................................................... 12.4 Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. ii License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 t

REVISION RECORD REVISION - ' DATE OF ' PAGES NUMBER REVISION REVISED REVISION REASON 1.0 - 30APR95 All Update to current operations. 2.0' 28JUN% iii, 6.8 Clarify Criticality Safety Basis for the compaction operation. 3.0 30AUG96 iii,1.7,1.9,12.6,12.7 1 Incorporate Safety Condition S-3 into Application; correct reference to Figure 1.3 instead of. 2.3, to reflect expansion of the CAA in order to eliminate need for gate. 4.0 30SEP96 iii, 6.11, 6.12 Clarification of Criticality Safety Basis for the Pellet Stripping System Equipment { and Hoods & Containment. 5.0 08NOV% lii,1.12,3.18, and 3.19 Incorporation of a dermition, (Reprinted all document - arxiincident notification pages in Microsoft Word criteria, recently approved format) by NRC Staff.- j ~6.0 05MAY97 6.12 (Reprinted all Clarify Evaluation document pages in Bounding Assumptions ' Microsoft Word format.) for Storage of Annular Pellets. 7.0 14JUL97 iii,12.2 and 12.3. Withdraw an existing. o authorization, and expand another ' authorization to enable cement manufacturing with CaF - 2 ' 8.0 11AUG97 iii,2.4 and 8.1 (Reprinted Change emergency exercise all document pages in frequencies for consistency l Microsoft Word format.) with Emergency Plan. Docket No. 70-1151-Initial Submittal Date: 30APR90 Page No. iii h i License No. SNM-1107 - Revision Submittal Date: 23JUL98 Revision No. 14.0 l

O 9.0 23SEP97 iv Chapter 6 To respond to NRC Staff request for additional information. To revise table to correlate to CSE organization and clarify discussions regarding margin of safety with respect to normal operations, expected process upsets and credible process upsets. 10.0 31 MAR 98 Table of Contents i-iv To replace Revisions Chapters 3.0 and 6.0. Numbers 2.0, 4.0, 6.0, and 9.0; and respond to SMIP initiative regarding SNM-l 1107, Chapter 6.0. (Chapter j 3.0 shown with bars & 6.0 l Major Rewrite). 11.0 03APR98 Table of Contents i-iv, To reflect common 1.12, 3.18 - 3.20. understanding on notification criteria with NRC Staff. i 12.0 30JUN98 Table of Contents (iv), To update and enhance Chapter 1.0 (1.12), Integrated Safety Assessment Chapter 4.0 (all). commitments. 13.0 13JUL98 Table of Contents (iv). To update and enhance .{ License Safety Condition S-2. 14.0 23JUL98 Table of Contents (iv). Chapter 1.0 (1.4,1.5); Tc reflect current Chapter 2.0 (2.0, 2.1). organization Chapter 2.0 (2.3, 2.5), To clarify commitment to Chapter 3.0 (3.7, 3.8, 3.9). . integrated safety. Chapter 3.0 (3.3). To delete _ pellet caits from programmed maintenance. Chapter 3.0 (3.15). To carify commitment to formal audits. Chapter 5.0 (5.9, 5.10). To expand commitment for invivo bioassay, i Docket No. 70-1151-Initial Submittal Date: 30APR90 Page No. iv f License No. SNM-1107 . Revision Submittal Date: 23JUL98 Revision No. 14.0 \\ = _ - - _ _ _

) Chapter 5.0 (5.11). To clarify commitment to respiratory protection. Chapter 8.0 (8.9). To update pre-fire plan preparation to current practice. I J l 1 l l l 'I l LDocket No. 70-1151 Initial Submittal Date: 30APR90 Page No. _v License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l l

CHAPTER 1.0 GENERAL INFORMATION l' L 1.1 FACILITY AND PROCESS DESCRIPTION The Columbia Fuel Fabrication Facility (CFFF) of the Commercial Nuclear Fuel l' Division (CNFD) will be primarily engaged in the manufacture of fuel assemblies for commercial nuclear reactors. The manufacturing operations to be authorized by this license will consist of receiving low-enriched, less than or equal to 5.0 w/o U-235, L uranium hexafluoride; converting the hexafluoride to produce uranium dioxide powder; l. and processing the uranium dioxide through pellet pressing and sintering, fuel rod loading and sealing, and fuel assembly fabrication. These operations will be governed by the L technically sound radiation and environmental protection, nuclear criticality safety, L industrial safety and health, SNM safeguards, and quality assurance controls described in detail in this License Application. i-Two' general systems are used to convert uranium hexafluoride to uranium dioxide powder-Integrated Dry Route. (IDR) and Ammonium Diuranate (ADU). IDR conversion equipment has been designed to receive and process uranium in enrichments up to 5.0 w/o U-235, through fuel rod loading. ADU conversion equipment has also been designed to receive and process uranium in enrichments up to 5.0 w/o U-235, ' through fuel assembly fabrication and shipping. These operations are supported by l absorber coating, laboratory, scrap recovery, and waste disposal systems. Additional - details concerning the facility and process systems are presented in the Site Safeguards documents described in Paragraph 1.1.1(e) of this Section, and in the SITE L EMERGENCY PLAN described in Chapter 9.0 of this License Application. 1.1.1 SITE UTILITIES AND SERVICES u (a) Electrical Supply The CFFF will be served by a single, 115,000 volt, electrical supply line. Four l ' diesel-powered standby generators will be installed and maintained to meet the y emergency electrical power requirements of the site in the event of a temporary outage of the normal supply source. Emergency power will be automatically provided to cmcial process equipment; emergency lighting systems; cooling system pumps; all fire alarm, hazard alarm, and other designated safety alarm systems; Conversion Control Room alarms; health physics sampling systems; and, emergency ventilation systems, including scrubbers. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.0 l l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No.14.0 l i

l (b)- Water Supply A ten-inch main from the Columbia Municipal Water Authority supplies water to the site. (c) Gaseous and Liquid Effluent Management Gaseous exhausts, with potential for contamination, from process areas will be routed through HEPA filtration, to remove entrained uranium particulate, prior to discharge to the environment. Exhausts containing uranium in soluble form will be passed through aqueous ~ scrubbers, preceding the HEPA filters. . Following filtration, the gases will be continuously sampled, to enable analyses - for esuring compliance with the limits specified in this License Application. Liquid process wastes will be treated, prior to discharge to the Congaree River. Waste treatment, for the. removal of uranium, ammonia, and fluorides, will 3 consist of filtration, flocculation, lime addition, distillation, and precipitation (in a l series of holding lagoons). Site sanitary sewage will be treated in an extended aeration package plant prior to discharge, either directly or through a polishing lagoon. The discharged effluent will be chlorinated, and mixed with treated i liquid process waste, at the facility lift station. The combined waste will then be passed through a final aerater, followed by pH adjustment as required, and subsequently pumped to the river via a 4-inch pipeline. Compliance with licensed ) limits will be verified by passing the waste streams through on-line monitoring systems, or by manual sampling and analysis on a batch-basis. The treatment systems will have sufficient holdup capacity to assure the limits are continuously met. Storm water from the site enters a system of drainage ditches and ultimately flows to the Congaree River. (d) SOLID WASTE STORAGE AND DISPOSAL Solid wastes will be sorted into appropriate combustible and noncombustible fractions, and placed in specially designated collection containers located j throughout the work area. (The wastes consist of paper, wood, plastics, metals, floor sweepings, and similar materials which are contaminated by, or contain, uranium.) Following a determination that the wastes are in fact properly sorted, I the contents will be transferred to a waste processing station. Materials that are suited for thorough survey may be decontaminated for free-release, or re-use, in accordance with provisions of this License Application. Combustible wastes will be packaged in compatible containers, assayed for grams Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.1 l l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

r U-235, and stored to await incineration. Noncombustible wastes, and selected combustible wastes, will be packaged in compatible containers, compacted when appropriate, measured to verify the uranium content, and placed in storage to J await shipment for further treatment, recovery, or disposal. j Administrative controls will be in effect to assure that only authorized materials are packaged for disposal. (These include verification of package contents, container security to minimize the probability of unauthorized additions to the - containers, documentation of package contents, and routine overchecks to verify l that the above referenced controis are effective.) Wastes designated for disposal l will be packaged in DOT approved 55-gallon metal drums or in metal boxes. Materials packaged in metal boxes will be pre-measured in standard containers prior to transfer to the boxes. Filled containers will be stored in designated areas L within the manufacturing or waste storage buildings; or, they may be stored f outdoors, if protected from the elements. Wastes consigned to disposal will be shipped to a licensed burial facility. Shipnants will be made in compliance with all applicable NRC, DOT and State regulations; and, in conformance to burial site criteria. (e) SITE SAFEGUARDS Nuclear Materials Control and Accounting at the CFFF is described in the NRC-approved FUNDAMENTAL NUCLEAR MATERIAL CONTROL PLAN FOR THE COLUMBIA FUEL FABRICATION FACILITY, dated April 1, 1987, and subsequently revised in accordance with the regulations. Physical Security at the CFFF is described in the NRC-approved PHYSICAL SECURITY PLAN FOR THE COLUMBIA FUEL FABRICATION FACILITY, dated September 1,1984, and subsequently revised in accordance with the regulations. These Plans detail the measures employed at the facility to detect any potential loss of, and mitigate the opportunity for theft of, Special Nuclear Material of Iew Strategic Significance, in accordance with applicable requirements of 10CFR73 and 74. . L1.2 SCOPE OF LICENSED ACTIVITIES Compliance with all applicable Parts of Title 10, Code of Federal Regulations will be required, unless specifically amended or exempted by NRC staff. (a).. Authorized Activities: (a.1). Authorized activities at the Columbia Fuel Fabrication Facility will include: (1) Receipt, handling, and storage of Special Nuclear Material a.c uranium Docket No. 1151 Initial Submittal Date: 30APR90 Page No. 1.2 l License No. SNM-1107 ' Revision Submittal Date: 23JUL98 Revision No. 14.0 ~l l

c hexafluoride, uranium nitrates, uranium oxid s; and/or contained in pellets, fuel rods, fuel assemblies, samples, scrap, and wastes; (2) Receipt, handling, and storage of other licensed radioactive material; (3) Chemical conversion processing by the Ammonium Diuranate Process and the Integrated Dry Route - including vaporization and hydrolysis, precipitation and centrifugation, drying, calcining, comminution, and blending; (4) Fuel fabrication - including powder preparation, die-lubricant mixing, pelleting, sintering, grinding, pellet coating with ' nuclear absorbers, fuel rod loading and inspection, and final fuel assembly; (5) Quality assurance and control inspection activities; (6) Analytical Services Laboratory j operations - including wet-chemistry 'and spectrographic. techniques; (7) l Metallurgical II.boratory operations - including sample preparation, polishing, testing, and examination: (8) Chemical Process Development operations - including laboratory-scale process research, prototype development, and { equipment check-out; (9) Mechanical Process Development operations - q including laboratory-scale research and development; (10) Health Physics L Laboratory operations - including sample preparation and analysis, instrument repair and. calibration, respirator fit-testing, and bioassay sample and scaled-source storage; (11) In-house, and contracted, scrap recovery operations - including scrap batch processing, solvent extraction, coated-pellet recovery, scrap blending, and hydrofluoric acid recovery; (12) UF. cylinder washing, hydrostatic testing and re-certification; (13) Equipment and facility maintenance activities; (14) Equipment and facility decontamination activities - including clothing; (15) i Waste. storage and disposal preparation operations - including HEPA filter I

testing,; conversion liquid waste treatment, advanced waste-water treatment, lagoon storage, incineration, radioactive waste packaging for disposal, and calcium fluoride disposition; (16) Ancillary mechanical operations - including

. non-radioactive component fabrication and assembly; and (17} 3 hipping container and overpack refurbishment. (a.2) The licensed activity may also perform work for other Westinghouse Divisions, or outside customers, which is within the authorized capabilities of the facility. (b)- Material Possession Limits and Constraints 1 The following will be the maximum quantities of Special Nuclear Material that may be possessed by the licensed activity at any one time; and, constraints for procurenant, use, and transfer of such material. l .(b.1) Material possession limits - (1) 5-grams of U-233 in any chemical or physical [l form, lireited to !aboratory use as individual 1-gram maximum quantities in ventilated hoods; (2) 350-grams of U-235, as uranium of any enrichment, in any chemical or physical form; (3) 75,CM kilograms of U-235, as uranium enriched Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.3 l IM No. SNM-1107 Revision S bmittal Date: 23JUL98 Revision No. 14.0 l 1

to no greater than 5.0 weight-percent, in any chemical or physical form except metal; and, (4) 1.5-grams of Pu-238/239 as sealed sources. (b.2) Material constraints - The procurement of Special Nuci. ear Materials will be in accordance with licensed activity needs. -Production, utilization, and/or significant loss of special nuclear tnaterials will not be authorized. Transfers of Special Nuclear Materials will be only as arranged with facilities authorized to receive and possess such materials. 1.2 INSTITUTIONALINFORMATION This application requests a ten year renewal of License SNM-1107, Docket 70-1151, which authorizes the receipt, possession, storage, use, and transfer of Special Nuclear Material at the Westinghouse Electric Company's Columbia Fuel Fabrication Facility l (CFFF). There is no control or ownership exercised over Westinghouse Electric Company by any alien, foreign corporation, or foreign government. In accordance with l the requirements of 10 CFR 70.22(a)(1), the following additional information is submitted: '1.2.1 APPLICANT AND STATE OF INCORPORATION ' WestingNw Electric Company l Pennsylvania 1.2.2 LOCATION OF THE PRINCIPAL OFFICE Pittsburgh, Pennsylvania 1.2.3 NAMES (CITIZENSHIP) AND ADDRESSES OF PRINCIPAL OFFICERS Charles W. Pryor (USA) President and Chief Executive Omcer Westinghouse Electric Company Westinghouse Energy Center l P. O. Box 355 Pittsburgh, Pennsylvania 15230-0355 Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.4 l l l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

James A. Fici (USA) l General Manager, Commercial Nuclear Fuel Division Westinghouse Energy Center P. O. Box 355 Pittsburgh, Pennsylvania 15230-0355 Jack B. Allen (USA) l CFFF Plant Manager Westinghouse Columbia Plant Drawer R Columbia, South Carolina 29250 1.2.4. CORPORATE CONTACT FOR LICENSING MATTERS ) . Griff Holmes Manager, ESBU EHS Regulatory Affairs l -Westingk=m Energy. Center ) P. O Box 355 Pittsburgh, Pennsylvania 152304355 1.2.5 SITE CONTACT FOR LICENSING MATTERS I Robert A. Williams l

Licensing Project Manager l

Westinghouse Columbia Plant Drawer R Columbia, South Carolina 29250 1.2.6 ADDITIONALINFORMATION Additional corporate financial and business information is provided in the Westinghouse Annual Report, available from: Westinghouse Electric Company l P. O. Box 8815 Pittsburgh, Pennsylvania 15221 l 1.3 SITE DESCRIPTION I The Columbia Fuel Fabrication Facility (CFFF) is located near Columbia, South Carolina and is situated on an approximately 1,158 acre site in Richland County, some' 8 miles l' southeast of the city limits of Columbia (see Figures 1.1 and 1.2) along South Carolina ) I Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.5 l l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l J

~ { i FIGURE 1.1 CFFF SURROUNDING AREA N W- -E 1 SOUT,H,,,CA,ROL. INA l e-One inch equals approntmately 35.6 miles W Y), n.1 21 cw gesceu ,~"l ._ 7 ,'..,, - &,_$t g 4..s_ ~ ~ c. ) -aaf j i 5" p* / \\ ~\\'. weassare n*,5; "u .. 3 A =~- (\\ h \\5 N 8 G==ai-m i. -s ff,., - h ./, -&q- ~ h, la 33 erhesuu

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(,1,<alr AN"$~; ,AV..' 3,,,.,,( rrep j s b( ( \\# 3 ~E ^i UU0iialA ~ u u 31 16 2 r. .u- ,g,,x A w.ds.6.u r*' j w im eram,% c,t, a 29 F1 * Watores l,9 33 34 W i ss 1h W l "^'I'"'" .h d"""" gy n \\ / =tig,,, a l m 5 ~-}x ;D#"'"* j% m, y .. q,.... - u , g,,,4.,.,,,... p.me* Y',*,,,,,,,,,, a g,7 dw l 9 *ay .,x .i. l 1 i,, m. ',,,n,4 4.. y ,,., c 5,.,,,=,,,,, t g E :! %, cr;j ",, <=h8 ar-*=. .s".J i n. c., 3 l i Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.6 l f License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l ._w

i FIGURE 1.2 { CFFF PROPERTY BOUNDARY l l l (s.c. at. so) n. \\ \\ p I i f*I { f. \\u -,, q,- ,J \\/ \\ l l + ff i b , s f vp \\ Y \\ o u ,e a I / s c' o l'l ') %;+s Jp Q -jb@ e l \\ l CHF-92002 SHT-1 "' See Figure 1.3 Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.7 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

n l Highway 48. The region around the site is sparsely settled, and the land is characterized f l by timbered tracts and swampy, areas, penetrated by unimproved roads.

Farms,

{ single-family dwellings, and light commercial activities are located chiefly along nearby highways. i l The site is bordered by abutting properties, as presented in.the PHYSICAL SECURITY ). PLAN described in Paragraph L1.1(c) of this License Application. Approximately 1098 !~ acres of the site remain undeveloped. Of the total 1,158 acres, only 60 acres (about 5 percent) have been developed to accommodate the fuel fabrication facilities, holding ponds, and !=ieapM areas. A site plan is shown in Figure 1.3. Details of the CFFF location, meludmg proximity to nearby towns, industries, public - facilities, the Congaree River, transportation links; and, site topography; are presented in Section 1 of the SITE EMERGENCY PLAN. Details of the site characterization are presented in Section 2.0 of the SITE EVALUATION REPORT. 1.4 TERMS AND DEFINITIONS Throughout this License, the following terms will be defm' ed and used as indicated: ALTERNATIVE ACTIONS - Tests, procedures or other practices that may be l substituted for prescribed activities as deemed appropriate by the Regulatory Component. In such case, a detailed analysis will be performed and documented by the cognizant

Regulatory Functions. This analysis will include a comparison of the proposed action with that specified in the license; and, a demonstration that action hvels and limits of the license will be met, and that health and safety of employees and the public, and quality of the environment, will be protected.

CHEMICAL AREA - An area where uncontained radioactive material is processed, the probability of contamination on floors and accessible surfaces.is high, and protective clothing is required; such as, the UF, Bay, the Conversion Area, the Pelleting Area, the Rod Loading Area, etc. CLEAN AREA - An area where radioactive material, if present, is completely contained and there is negligible contamination on the floors or accessible surfaces. Such locations include, but are not limited to, the Machining Area, Grid Assembly Area, Final l Assembly Area, Office Areas, and the Cafeteria. COMPONENT - When used in an administrative context, an independent organizational unit distinguishable by its assigned responsibilities; such as, the Engineering Component, the Manufacturing Component, the Quality Component, and the Regulatory Component.- Docket No. '70-1151 Initial Submittal Date: 30APR90 Page No. 1.8 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

FIGURE 1.3 SITE PLAN e 2= e e ~ 0 ~ ca. e l Ao e 31, m a u n n /. g M' e P3 a e ' O ,, fg T" e O 'Ha"" r- ._/.* 1X o + vas

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.. / o e e ErP E. I e g 1, LEGEND: 8 EELl (3 A NED caer e @ E. Y '*rvzMS rotwrcentarTv cc-teoon sur-e rw.os Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.9 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l L

1 1 l 1 SITE PLAN KEY S1 : Shipping / Receiving point for UF. Cylinders. R1 S2 : Shipping / Receiving point for Uranyl Nitrate. R2 S3 : Shipping / Receiving point for U Powders, U Pellets, R3: U Scrap, and U Waste. S4 : Shipping / Receiving point for U Rods and U Assemblies. j R4 I l S5 : Shipping / Receiving point for miscellaneous U Samples R5 and U Standards. I: Fuel Manu'acturing Building. 2: UF. Storage Pad. 3: Gatehouses (4). { i 4: Administration Building. I 5: Parking Area. 6: L.agoons (6). 7: Waste Treatment Building. 8: Waste Storage Area. 9: Access Road. 10 : Controlled Access Area (CAA) Fence, 11 : Advanced Liquid Waste Treatment Building. 12 : Uranyl Nitrate Storage Tanks. 13 : Fuel Manufacturing Building Southwest Expansion. 14 : Fuel Manufacturing Building Southeast Expansion. 15 : Nuclear-Poisoned Fuel (IFBA) Area. Docket No. __7_0-1151 Initial Submittal Date: 30APR90 Page No. 1.10 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l 4 J

L I CONTAMINATION CONTROLLED AREA - An alternate mune for the Chemical Area. CONTROLLED ACCESS AREA - A physically defined area, represented on three sides by a seven-foot high barner of Number-11 American Wire Gauge fabric-fence topped by three strands on barbed wire, and on the fourth side by the Admmistration and Main Manufacturing Building. This area is the " Controlled Access Area" described in the Physical Security Plan. l ENRICHMENT LIMIT - When used as an authorized enrichment limit, 5.0 w/o U-235 means that, based on an enrichment measurement uncertainty no greater than 0.50 percent relative, the hypothesis that the tme enrichment level is 5.0 w/o U-235 or less can not be rejected at the 0.05 level of significance. 1 EQUIVALENT EXPERIENCE - When used in a personnel qualification context to equate experience with education, eight years of applicable expenence is equivalent to a j baccalaureate degree. FIXED LOCATION GENERAL AIR SAMPLE - Air samples used to assess general . area radioactivity concentrations; and, to assess the adequacy of radioactive material containment and confinement within the processing areas of the facility; and, to establish I airborne radioactivity areas. FIXED LOCATION BREATHING ZONE REPRESENTATIVE AIR SAMPLE - Air samples used for purposes of assessing and assigning operator intake. FREQUENCIES - When measurement, surveillance,' and/or other frequencies are specified in License documents, the following will apply: DAILY means once each 24-hour period; WEEKLY means once each seven consecutive d MONTHLY means twelve per year, with each covering a span of 404ays or less; TERLY means four per year, with each covering a span of 115-days or less; SEM NUAL means two year, with each covering a span of 225-days or less; ANNUAL means once per year, per not to exceed a span of 15-months; BIENNIAL means once every two years, with each covering a, span of 30-months or less. TRIENNIAL means once every three years, with each covering a span of 45 months or less. FUNCTION - When used in an administrative context, an individual (or individuals), designated by the Component Manager, acting in coordination with the other personnel of a component, having the capability, responsibility ' decisions required to carry out assigned duties;y, and authority to make and.. imple such as the Environmental Protection ' Function, Radiation Safety Function, Nuclear Criticality Safety Function, Chemical Safety Function,. Fire Safety Function, and Sat: guards Function of the Regulatory Component. LICENSED ACTIVITY - That combination of personnel, plant, and equ established. by Westinghouse Electric Corporation to carry out the processm,ipme radioactive material authorized by this License Application. J MAY -- Denotes implied permission by NRC Licensing Staff to take a stated action or course. 1 PORTABIE ~ AIR SAMPLE - An air sample that is not integrated into the plant's I central air sample vacuum system. l 1 i L l . Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.11 l License No. ' SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l 1

REGULATORY-SIGNIFICANT PROCEDURES - Those procedures that contain, in whole or in part, actions that are important to environmental protection, health, safety, and/or safeguards. RESTRICTED AREA - Areas such as the Manufacturing Building, or equivalent areas, to which access -is restricted physical or administrative methods and which is . monitored on a scheduled basis the site Security Function. SAFETY MARGIN IMPROVEMENT CONTROLS - Controls that effective enhancements to the safe and effective, operation of a process. provide These are controls that enhance an existing and adequate margm of safety. SAFE MASS [3.7.3(b.2) and (c.5)] critical mass for a particular process or vessel given the' credible material geometry for that process / vessel, and the License Evaluation Bounding Assumptions for that material type (e.g., homogeneous UO ) and reflection. 2 Optimum moderation and material density are assumed. SAFETY-RELATED - Relevan: to systems crucial or important to safety; and, those systems that improve the margin of safety (e.g., in the context of maintenance). SAFETY-RELATED CONTROLS - Preventive and mitigative controls relied upon for environmental protection, radiation safety, nuclear criticality safety and safeguards, I chemical safety, and fire safety. These controls, which include both " Safety-Significant" I and " Safety Margin Improvement" controls as sub-sets, will be identified through an integrated safety analysis which documents the design safety basis for a particular process. SAFETY-SIGNIFICANT - Relevant to systems crucial or important to safety (e.g., in the context of quality assurance). SAFETY-SIGNIFICANT CONTROLS - Controls crucial or important to, or deemed desirable for, the safe and effective operation of a proces.s, and an adequate safety margin l for the process. An adequate safety margin is made up of those controls necessary for ) compl,fe operation of the process plus those controls identified to ensure regulat the sa iance. UNRESTRICTED AREA - An area, access to which is neither limited nor controlled. WILL - Denotes a mandatory requirement to take a stated action or course. J l l l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 1.12 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l [~ ___i

R CHAPTER 2.0 MANAGEMENT ORGANIZATION 2.1' ORGANIZATIONALRESPONSIBILITIES AND AUTHORITIES The Westinghouse Electric Company (WEC) is divided into business units. One such l unit is the Energy Systems Business Unit (ESBU), which encompasses all of the commercial activities of WEC directly related to the development,' manufacturing, and marketing of products contributing to the use of nuclear reactors for electrical power generation. 2.1.1 ORGANIZATIONAL OPERATING UNITS ' Within ESBU, the primary responsibility for the design, development, and manufacture of commercial nuclear reactor fuel rests with the Commercial Nuclear Fuel Division (CNFD). The General Manager of CNFD reports directly to the Vice President and General Manager of ESBU, Within CNFD, the prunary responsibility for all commercial nuclear reactor fuel manufacturing activities rests with the Columbia Fuel Fabrication Facility (CFFF); the CFFF Plant Manager reports to the General Manager of CNFD. Figure 2-1 illustrates the general structure of the Corporate organization. The ultimate responsibility for all CFFF activities associated with the manufacture of - commercial nuclear reactor fuel -- including environmental protection, health, safety, i quality, ad safeguards. rests with the Plant Manager. The site organization consists of several staff Components reporting directly to the Plant Manager. One of these Components, Regulatory, has the responsibility for overall coordination and . implementation of the Columbia Plant environmental protection, health, safety, and -l safeguards programs. Figure 2-2 illustrates the general structure of the CFFF organization.- .I i 2.1.2D POSITIONS AND ACTIVITIES WITHIN ORGANIZATIONAL OPERATING UNITS Each Westinghouse management position is covered by a written description, presenting in detail its scope, purpose, dutics, responsibilities, difficulties, and requirements. The . description identifies' the incumbent's authority for decisions which may be made E .' unilaterally, and those requiring higher management approval. It delineates relationships L with other functions, and specifies responsibilities for managing personnel, and for the i control and maintenance of managed facilities and equipment. Position descriptions are } reviewed and approved by two higher levels ofline management. A Management Docket No. -_70-1151 Initial Submittal Date: _30APR90 Page No. 2.0 l License No. SNM-1107 - Revision Submittal Date: 23JUL98 Revision No. 14.0 l l t___--___ i

T FIGURE 2-1 CORPORATE ORGANIZATION WESTINGHOUSE ELECTRIC COMPANY l (WEC) l l l ENERGY SYSTEMS BUSINESS UNIT (ESBU) COMMERCIAL NUCLEAR FUEL DIVISION (CNFD) l COLUMBIA FUEL FABRICATION FACILITY (CFFF) i l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.1 l l 1 License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

l !^ FIGURE 2-2 CFFF ORGANIZATION t PLANT MANAGER ~ ENGINEERING MANUFACTURING REGULATORY QUALITY COMPONENT COMPONENT COMPONENT COMPONENT l "" "'En'cTo"u""* b d oremanons ruwerion l d assuaance runction l ' l raciurg,ge u.oy H mame=aace ruacno= l q ,,,,,crio,uncria l l a g g,g e.. l l 1 l j l --"va :"a"l l "%M'JT l-i *"*"fac'^=" l l

c

l u -^=a" - l l " = a = ar i l 1 1 l 1 i ? i Docket No. 70-1151-Initial Submittal Date: - 30APR90 Page No. 2.2 l [' License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

[- Position Committee, which consists of key members of the CNFD staff, reviews and . evaluates such positions. These reviews determine that all key functions are covered, 1 inter-relationships are clear, and conflicts are eliminated. Persons are selected to fill- ) these management positions by evaluating their capability to perform the various activities E specified in the position description. Two higher levels of management, at a minimum, i must approve each selection or change of a management incumbent. Continuing quality i performance of managers is assured through a formal program of annual review, i Operations at the Columbia Fuel Fabrication Facility are in accordance with the general l- ' operating philosophy and procedures that are employed in all Westinghouse plants and facilities. Briefly, this philosophy provides that total responsibility for all phases of operations, including environmental protection, health, integrated safety, quality, and l L safeguards follows the usual lines of organizational authority. Advisory and service groups are provided to assist line management in the analysis of operations within their cor. trol, and to provide measurements, determinations and information which aid in the -analysis of specific operations and situations; however, such service and staff assistance L in no way relieves an individual line manager' from accountability for high quality operation of the function and facility, or for-ascertaining and assuring, through appropriate management channels, that adequate service is provided. Basic policies and procedures are established by line management with the review and approval of cognizant staff groups; and, within the framework of these policies and procedures, the responsibility for making decisions at the operating level rests with the first level L manager. 'A first level manager has the oasic responsibility for operating controlled activities in a safe and pmdent manner. l First level managers are responsible for providing operating instructions for the guidance l and direction of subordinate personnel. Written procedures or manuals are prepared, which become the bases for performing specific operations. The first level manager cannot make unilateral changes in such written instructions, or in posted limits, without review and approval of cognizant staff groups. First level managers are also responsible for assuring that personnel under their jurisdiction receive adequate training. The Regulatory Component presents an orientation to new employees. Fundamental l: radiation safety mies and policies, use of protective clothing and personnel monitoring i devices, prevention of internal exposure, limiting exposure to external radiation, nuclear criticality safety, and plant emergency procedures are among the topics discussed. To acquaint the new employee'with basic regulations, selected parts of Title 10, Code of L Federal Regulations, are covered. Primary emphasis is placed upon 10 CFR Parts 19 and

20. The cognizant first level manager assigns an experienced employee the responsibility of indoctrinating and training a new employee in the proper procedures and precautions for performing each specific job. The first level manager then evaluates the progress of the new employee and gradually increases job assignments until complete requirements of e_

n . Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.3 l License No. SNM-1107 Revision Submitu : 7 ate: 23JUL98 Revision No. 14.0 l 1 -___l__-______-

the job description are fulfilled. Failure to achieve minimum performance requirements is cause for a change in assignment, or for release from employment. Periodic

reinforcement instruction is conducted, on the job, by the employee's first level manager and/or by personnel from the Regulatory Component. As the need arises, changes in regulations, changes in operating conditions and/or procedures, and changes in administrative policies are covered.

To assure that all employees, who are not members of the emergency response organization, are aware of actions to take during an emergency situation, biennial training is provided. To keep emergency response personnel aware of actions they must take during an emergency situation, emergency drills and exercises are conducted in alternate years. After each drill or exercise is evaluated, appropriate first level managers are informed of any shortcomings disclosed, and they subsequently instruct their personnel regarding any remedial actions required. 1 At the CFFF, all personnel involved in operation of the facility will have the right to question, and/or request review of, the safety of any operating step or procedure. Further, a cognizant. Regulatory Component staff member.on duty will have the responsibility and c.athority to prohibit, through the cognizant first level manager, any operation which is believed to involve undue immediate hazard. Such terminated operations will remain in safe-shutdown until the situation is reviewed with cognizant management, and there is a consensus resolution of the methods and procedures to be used. 2.1.3 POSITION ACCOUNTABILITY AND REQUIREMENTS ') Administrative and managerial controls will be in effect at all times to assure that decisions related to the operation of the licensed activity are made at the designated level of accountability, by individuals meeting the necessary technical requirements. 1 (a) Plant Manager The Plant Manager will have overall accountability for all nuclear fuel I manufacturing activities at the Columbia Fuel Fabrication Facility. This l individual will' direct all activities of licensed operations and staff functions, either personally or through designated management personnel. This individual will also coordinate any 'necessary support activities, obtained from higher -Westinghouse management; and, will perform all assigned management functions in accordance with Westinghouse policies and higher management directives. .The minimum requirements for the position of Plant Manager will be a baccalaureate degree, or equivalent: and, five years of management experience in i 1 i Docket No. 70-1151-Initial Submittal Date: 30APR90 Page No. 2.4 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l l

- - - - - --- ] q I o a nuclear facility. The Plant Manager will have broad general knowledge I concerning the regulatory aspects of policies and procedures in effect at the Columbia Fuel Fabrication Facility. (b). Component Managers i Four Component Managers will have specific accountability for engineering, manufacturing, regulatory, and quality operations and activities involving licensed l-materials. The Manufacturing Component will conduct the operations and maintenance activities required for production of nuclear fuel. The Engineering ) Component will provide design services related to processes and facilities used by -I L the Manufacturing Component. The Quality Component will provide assurance, inspection, and analytical services in support of the Manufacturing Component. (The Regulatory Component is described in Paragraph (c) of this subsection.) Component Managers will plan, direct, and control such activities personally, or through other management personnel; and, will perform all assigned management. duties in accordance with Westinghouse policy and higher management directives. A Component Manager may be responsible for more than a single work area; and, will be directly accountable for the safe operation and control of activities in the work area (s) and for the protection of the environment, as influenced by the activities conducted. With appropriate support from cognizant service groups, they will be responsible for environmental protection, health, integrated safety, l quality, and safeguards, in all areas over which they have authority.. First level Managers will supervise operating personnel. They will fulfill their i responsibilities by assuring that all operations under their control are carried out l in accordance with the radiation protection limits, nuclear criticality safety ) controls, processing procedures, schedules, and other instmetions supplied by l higher management. l All Component Managers will be knowledgeable in the operating procedures i applicable to their work areas. Each Manager will have demonstrated proficiency in application of the licensed activity's environmental and radiological protection programs, as they relate to controls and limitations on work activities, in assigned radiation and radioactive materials areas. Each Manager of work areas where uranium is handled will have demonstrated proficiency in the application of the areas'. nuclear criticality safety controls. All Managers will be knowledgeable in the occupational safety and health procedures applicable to their areas of responsibility, i The minimum requirements for a Position of Component Manager, above the First level, will be a baccalaureate degree, or equivalent, with a science or s l - Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.5 l ~ License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

engineering emphasis; and, two years of experience in a nuclear facility. A First level Manager will have demonstrated management capabilities by a continuing record of quality work accomplishments. (c) Regulatory Component Managers and Engineering Functions The Regulatory Component will be that organizational component of the licensed activity with the responsibility for environmental pollution control, radiation protection, nuclear criticality safety, occupational safety and health, and 'anergency planning; and, for evaluating the effectiveness of these programs. The Regulatory Component will be specifically responsible for assuring that applicable license conditions, radiation and environmental protection requirements, nuclear criticality safety requirements, and occupational safety and health requirements i ave been evaluated and communicated to other Component management for incorporation into facilities, equipment, and procedures prior to their use for p.ocessing licensed material. ) Tha Regulatory Component will, to the extent practicable, be administratively indoendent of manufacturing process supervision. The Regulatory Component will be responsible'for the establishment, conduct, and continuing evaluation of licenscJ programs to ensure the protection of the employees at the licensed facility, or the publi::, ?.=1 cf tio environment. In particular, for any processing i change which could result in a credible consequence not previously evaluated, or in excess of one previously evaluated, the Regulatory Component will perform a L safety analysis to assure that no off-site consequences, in excess of those specified in the regulations, would occur. Any process change for which the analysis indicates that a process upset could produce effects in excess of those previously evaluated will be submitted for review and approval by the NRC staff, prior to implementation. The radiation protection program administered by the Regulatory Component will include'as a minimum: the evaluation of releases of radioactive effluents and j materials from the site; the establishment of procedures to control contamination, exposure of individuals to radiation, and integrity and reliability of radiation detection instruments; the maintenance of required records and reports to L document the program's activities; and a program to maintain the' above parameters As Low As Reasonably Achievable (ALARA). Nuclear criticality safety services provided by the Regulatory Component will l. include, as a minimum: the-performance of process or equipment nuclear i criticality safety analyses and evaluations before a new or modified fissile material ~ operation is begun to include the determination of parametric controls and spacing Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.6 l t. L License No. - SNM-1107 Revision' Submittal Date: 23JUL98 Revision No. 14.0 l 3 LL_-_- )

requirements based upon validated analytical or computational techniques, including computation of effective neutron multiplication factors for fuel configurations; provision of audits, inspection and surveillance services to protect against accidental criticality; the maintenance of required documentation for the program performance of process and equipment review, validated nuclear criticality safety analyses and evaluations, operating equipment and procedure review, verification, and approval; and the performance of audits of the nuclear criticality safety program. The occupational safety and health program administrated by the Regulatory Component will include as a minimum: the evaluation of potential physical, chemical, 'and fire hazards; the development and implementation of safety programs and. procedures designed to minimize accidents and injuries to l employees; the procurement and maintenance of industrial safety protection and ( monitoring equipment; and the maintenance of required records and reports to document the program's activities. Specific responsibilities of the Regulatory Component will include, but not necessarily be limited to, the followmg: j i License and permit administration; j Routine surveillance of operations; j Audits of licensed activities for compliance with applicable State and Federal regulations, licenses, and permits; and, documentation of these audits and actions, to facilitate corrective activities; Maintenance of the site regulatory plans; Maintenance of the site regulatory manuals; Maintenance of the site regulatory procedures; Conduct and review of nuclear criticality safety analyses; Review and approval ' of all site procedures specifically related to environmental and radiation protection, nuclear criticality safety, occupational safety and health, and emergency planning; Review and approval of design drawings of equipment, and layouts, associated with the processing, handling, and storage of nuclear material; Inspection of installed equipment for conformance with radiation protection, nuclear criticality safety, and occupational safety and health requirements; and, docu:nentation of said conformance; Review of nuclear criticality safety, radiation protection, and occupational safety and. health aspects of changes to equipment and operations associated with the processing, handling, or storage of nuclear material; i Training in, and monitoring the training effectiveness of, environmental i protection, radiation safety, nuclear criticality safety, occupational safety L V Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.7 l l. License No SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

L and health, and emergency planning; and, Monitoring, and reporting the effectiveness, of the program to assure l radioactivity in effluents and radiation exposures.are kept As Low As Reasonably Achievable (ALARA). The minimum requirements for a position of Regulatory Component Manager will be a baccalaureate degree, with biological' science, physical science, or engineering emphasis; and, two years of experience in assignments involving regulatory activities. A Regulatory Manager will have appropriate demonstrated proficiency in health physics, nuclear criticality safety, and/or industrial safety and hygiene; and, in quality administration of functional programs being managed. 1 The minimum requirements for a position of Regulatory Function En;/.neer will be a baccalaureate degree, or equivalent,'with science or engineering emphasia; j and, two years of nuclear industry experience in the. assigned function. A j Regulatory Function Engineer will have demonstrated. proficiency in quality J administration of the assigned position programs. l I 1 -2.2. SAFETY COMMITTEES J The. Regulatory Compliance Committee (RCC) will be responsible ;for. overall coordination of all licensing, compliance, and regulatory health and safety matters; and, - for developing policies and procedures relating to the use and storage of nuclear materials. Special responsibilities of the RCC will include: ~ Review and ' assessment of radioactive material releases to unrestricted areas, internal and external radiation exposures, and unusual occurrences; Review and assessment of health and safety programs; Review and assessment of the ALARA program; Self-assessments of regulatory performance; Review of noncompliance items, and assurance of implementation of corrective actions; and, - Serving as the 10CFR21 Safety Review Committee. The Regulatory Compliance Committee will also function as a management advisory group to assure that operations are conducted in a manner that provides maximum { possible protection from injury to employees; and, to assure that employee health hazard concerns are adequately addressed. The Regulatory Compliance Committee will be chaired by the Plant Manager, or by an individual formally designated by the Plant Manager. RCC membership will consist of 1 l i hi Ddet No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.8 l ' License Noi SNM-1107 : Revision Submittal Date: 23JUL98 Revision No. 14.0 l ___-_______a

/. h the Manager of the Regulatory Component, and at least three other Component Managers l who are qualified to evaluate plant operations from a regulatory and safety standpoint. l The committee will convene at least quarterly on a routine basis;.and, following any

process upset or procedural deficiency identified by the Regulatory Component for committee involvement, or when otherwise warranted by instant circumstances. The committee's fmdings, conclusions and recommendations will be formally documented to the Phnt Manager, following each meeting, and tracked in the meeting minutes.

Appropriate action will be taken, as required, to maintain and demonstrate. compliance with regulatory and ALARA requirements. The ' Regulatory Compliance Committee may formally delegate any part of its responsibilities, or assign specified projects, to qualified individuals or sub-committees. Reports of progress, and findings and recommendations, by such individuals or sub-committees will be formally submitted to the RCC for review at scheduled meetings. L l' I j l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 2.9, l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

CHAPTER 3.0 CONDUCT OF OPERATIONS The basis for total quality conduct of operations at the Columbia Fuel Fabrication Facility (CFFF) will be the Safety Margin Improvement Program (SMIP). This program will _be a . structured oversight process that maintains management awareness, and enables monitoring, of management-specified regulatory and process improvement activities; and, will be a management decision process for determining where and when resources will be allocated. This program will address, until their logical completion, elements of Environmental Protection Improvement, ' Criticality Safety Margin Improvement, Occupational Safety Improvement, and General Plant ~ Improven.ut. A responsible individual will be assigned accountability for each SMIP element initiative. The Safety Margin Improvement Program will not be a commitment tracking system; SMIP commitments will be followed to management-approved completion by the responsible I individual specifically assigned accountability for each particular initiative. This program will be a documented demonstration of CFFF Managements' strong commitment to evaluate, on a continuing basis, opportunities to improve the Plant margin of safety - with the understanding that: addition, change, and/or deletion of program elements and/or initiatives; continuation of . ongoing program elements and/or initiatives; and/or, additions, deletions and/or changes of i program implementation schedules - relevant to the Safety Margin Improvement Program - will always be at the discretion of the Plant Manager, as advised by the Engineering, Manufacturing, and Regulatory Components. l 3.1 CONFIGURATION MANAGEMENT To assure that design changes will not adversely impact on environmental protection, health, safety, quality, and/or safeguards programs at the Coltunbia Fuel Fabrication Facility (CFFF), a formal review process will be established to analyze new systems and i components, or modifications to existing systems and components, in order to reliably predict performance under normal operating conditions and potential process upsets. Stmetured hazard analyses, as conducted in accordance with Chapter 4.0 of this License l Application will specifically include analysis of verified drawings under configuration management. I 3.1.1 CONFIGURATION MANAGEMENT PROGRAM AND PROCEDURE j r The CFFF Configuration Management Program will embrace an approved procedure for . implementation cf proposed additions or changes to facility systems. The procedure will define the review and approval process to assure the impacted systems will continue to meet or exceed regulatory specification requirements of baseline safety assessments. The

Docket No.

70-1151 Initial Submittal Date: 30APR90 Page No. 3.0 l License No. ' SNM-1107 ' Revision Submittal Date: 23JUL98 Revision No. 14.0 l l t-

- procedure will specify documentation required to maintain a current record of existing system conditions. 3.1.2 CONFIGURATION MANAGEMENTIMPLEMENTATION The Configuration Management Program will be a major sub-element of the Safety Margin -Improvement Program described in the introduction to this Chapter. Configuration management will not be a substitute for procedures described in Subsection 3.4.1 of this Chapter, but will facilitate continuing compliance with their requirements - through responsible facility addition and/or change project reviews. 3.1.3 CONFIGURATION MANAGEMENT PROCESS The following sequence of activities will be utilized for all facility addition and/or change projects. Complexity of each project, and the issues involved, will determine the magnitude of effort afforded to each activity. (a) A project will.be formally opened for review by an assigned responsible individual completing a configuration change control form, and enclosing specified project information for the review process. ) '(b) Manufacturing, Engineering, And Quality Component Reviews Designated Manufacturing, Engineering, and/or Quality Component Functions will review the project proposal for economics, practicality, and technical merit. Formal approvals will be documented as part of the review package. (c) Regulatory Component Reviews For Approval Extent and depth of regulatory review of the project will be formally determined by an assigned Regulatory Component Manager. Designated Regulatory Component Functions will review the project proposal for impact on environmental protection, health, safety, and/or safeguards programs; and, for I compliance with applicable regulatory requirements and conformance to . regulatory commitments. Formal approvals will be documented as part of the i review package. L _(d)' Ancillary Programs and Procedures 1 Ancillary programs and procedures will be activated commensurate with identification of environmental protection, health, safety, and/or safeguards issues. Such programs will range from simple design reviews by cognizant multi- - Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.1 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

discipline Functions, through stmctured - What-If/ Checklist or Hazards and Operability Analyses. Formal approvals will be documented as part of the review package. .The Regulatory Component may issue conditional, documented approvals for preliminary and/or detailed project designs as the process advances. (e) . Specific documents to be updated will be formally identified as the process advances. (f) Drawings that are generated, or modified, will be maintained in a "For Construction" state until applicable installation is completed. Following installation, the "As Built" conditions will be recorded as " Released" drawings that represent actual system configuration. (g)' A project will be formally closed by the assigned responsible individual signing the configuration change control form, attesting that all required documentation has been updated, all required training has been completed, and the project has been terminated. 3.2 MAINTENANCE ' The purpose ~of the maintenance program for safety-related systems and components at - .. the Columbia Fuel Fabrication Facility (CFFF) will be to assure that this equipment is kept in a condition of readiness such that it is likely to perform its desired function when called upon to do so. The maintenance program will embrace three functional activities: Prop.wousi Maintenance, to include spudfied frequency calibrations; Periodic Functional Testing; and, Repair or Replacement, for systems and components that fail to perform to required standards. '3.2.1 PROGRAMMED MAINTENANCE OF SAFETY-RELATED SYSTEMS AND COMPONENTS The Manufacturing Component will utilize a suite of maintenance planning and control L computer programs to initiate work orders for programmed maintenance, and to record details of the execution of the work orders. The computer programs will include procedures for programmed maintenance of safety-related systems and components - prepared, reviewed, and approved in accordance with Subsection 3.4.1 of this Chapter. { 3 The following safety-related systems and components will receive programmed maintenance: Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.2 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No.14.0 l L__=_____-_-___.

Air Compressors;. - Emergency Electrical Generators; Fire Detection and Fire Control; e-Natural Gas Valves; ' Nuclear Criticality Detection; e Pressure Relief Valves; l Steam Boilers. l L 1 - Additional safety-related systems and components will be placed under programmed maintenance, as disclosed by the results of Integrated Safety Assessments described'in 1 Chapter 4.0 of this License Application. Until a system's Integrated Safety Assessment (ISA) is completed, safety-significant controls identified through enhancements of the system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or new controls identifkxi through configuration management reviews of modifications of the system, will be scheduled, as necessary, for programmed maintenance to assure that } the controls are maintained at their original level of availability. Other specified safety- ) related controls for a system will be placed under programmed maintenance at the { disemtion of the cognizant Regulatory Engineer Function. Prograuuned maintenance of safety-related systems and components will include specified ( calibration and re-calibration of relevant instruments. Such calibration and re-calibration - will be' initiated and controlled by the maintenance' planning and control computer - programs. Discrimination between safety-related and non-safety-related calibrations will be by use of an entry on the electronic instrument calibration card utility within the maintenance planning and control computer programs. 3.2.2 PERIODIC FUNCTIONAL TESTING OF SAFETY-RELATED SYSTEMS AND COMPONENTS The. following safety-related systems and components will receive programmed maintenance at the frequencies indicated: Plant-wide Fire Alarm System and Criticality Alarm System - Each working shift, one day per working week; L Plant-wide Hazard Warning System - Semiannual; e Specified Safety-related Interlocks on Process Equipment - Annual; L 'e Hydrogen and Natural Gas Line leak Tests - Annual. 1 Additional safety-related systems and components will be placed under periodic l - functional testing, based on the results of integrated safety assessments described in m Chapter 4.0 of this License Application. Until a system's Integrated Safety Assessment (ISA) is completed, safety-significant controls identified through enhancements of the ) . Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.3 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l u_______

system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or i new controls identified through configuration management reviews of modifications of - the system, will be scheduled, as necessary, for periodic functional testing to assure that the controls are maintained at their original level of availability. Other specified ' safety-related controls for a system will be scheduled for periodic functional testing at the discretion of the cognizant Regulatory Engineer Function. 3.2.3 REPAIR OF SAFETY-RELATED SYSTEMS AND COMPONENTS The maintenance planning and control computer-generated work orders and records will provide documentation of systems and components that have been repaired or replaced. 1 When a component of a safety-related system is repaired or replaced, the component will be field-tested to assure that it is likely to perform its desired function when called upon to do so. If the performance of a repaired or replaced safety-related component could be different l from' that of the original component, the safety-related system will be field-tested to assure that it is likely to perform its desired function when called upon to do so. - 3.3 QUALITY ASSURANCE The purpose of the formal quality assurance _ (QA) program for safety-significant processing equipment at the Columbia Fuel Fabrication Facility (CFFF) will be to assure I that such equipment is designed, installed, operated, and maintained so.that it will l perform its desired function when called upon to do so. This quality assurance program will be in addition to the quality assurance programs for nuclear components and fuel shipping containers; however, the three programs may share common elements (e.g., , organization stmetures, tool and gage control, change management, etc.). 3.3.1 QA PROGRAM STRUCTURE To the maximum extent practicable, the.QA program for safety-significant processing equipment will utilize elements of the facility's Process Safety Management (PSM) l program (29 CFR 1910.119), structured to include licensed radioactive materials. The [ Engineering Component will maintain a detailed matrix that graphically demonstrates' j L how the PSM program elements will address the following QA program criteria: (a) QA Organization; (b) QA Program; 1 Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.4 l L i. License No.JSNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l p l

o, (c) . Equipment / System Design Control; (d) Procurement Documentation Control; t-(e) Instructions, Procedures, and Drawings; (f) Document Control; (g)(

Control of Purchased Materials, Equipment, and Services; (h)

Identification and Control of Materials, Parts, and Components; (i) Control of Special Processes; (j) : InternalInspections; (k) Test Control; (1) Control of Measuring and Test Equipment; (m) Handling, Storage, and Shipping Control.c; -(n) Inspection, Test, and Operating Status; - (o) Control of Nonconforming Materials, Parts, or Components; (p) . Corrective action; i (q)' QA Records; and, (r) ' Audits.' The PSM program will then be supplemented, as required, to assure detailed inclusion of all QA criteria. L 3.3.2. GRADED APPROACH The " graded approach" will be addressed by performing a systematic and integrated l assessment of the hazards at the facility; then, identifying the safety systems and [ components that are intende:I to prevent, or mitigue the consequences of, these hazards; then, to apply the programs of assurance which provide the appropriate level of quality. i Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.5 l j l ' License No. ' SNM-1107 Revision Submittal Date: 23JUL98 Revisior No. 14.0 l l i L___--_-___

l I l (Completion of these assessments, as an ancillary supporting process, will be phased-in according to the implementation schedule for the facility's Integrated Safety Assessment.) Where judgement is. required,. salient decisions will be documented; when quality requirements are determined not to be necessary, the bases will be documented. .(a) - Quality level A; Crucial Safety Systems These systems are crucial to safety and, therefore, will receive rigorous attention to installation, operation, and quality assurance. They will be dermed by controlling the following hazard consequences: Greater than or equal to 5 rem dose equivalent to an individual offsite; e and/or, Greater than or equal to 10 milligrams soluble Uranium intake by an individual offsite; and/or, Greater than cr equal to 25 milligrams HF/m' czposure to an individual-i e offsite. Crucial safety systems will require full application of the -QA program requirements, where each of the 18 criteria that could apply are specifically addressed. They will be initially qualified when placed into service, and will be requalified as required, using controlled methods and procedures. (b) Quality Level B; Impartant Safety Systems These systems are important to safety and, therefore, will include key aspects that require high quality judgement or attention to detail. The key aspects will be identified and documented in the hazard assessment. They will be defined by controlling the following hazard consequences: Greater than regulatory limits to an individual offsite; l Death or serious injury to an individual onsite. e L-Important safety systems will require selected application of the QA program requirements, where elements of the 18 criteria that the Quality Component determines will apply are specifically addressed. (c) Quality level C; Safety Margin Improvement Systems y These systems have safety implications, but are neither crucial nor important to safety. They do not require specified attention to quality assurance, and no l extraordinary level'of safety detail is applied. Safety margin improvement Docket No. ' 70-1151 - Initial Submittal Date: 30APR90 Page No. 3.6 l License No. SNM-1107: Revision' Submittal Date: 23JUL98 Revision No. 14.0 l j

.w systems will be maintained and operated as part of routine and prudent industry practice. t 3.3.3 : ADDITIONAL QA PROGRAM COMMITMENTS AND EXCLUSIONS The program will be designed and incorporated, as an ancillary supporting process of the facility's Integrated Safety Assessment, such that it becomes an integral part of routine . CFFF operations. The program will be performance-based. Quality assurance decisions will be based, to the extent practicable, on system performance histories. The program descriptions will be documented in facility procedures that specify responsibility, authority, and accountability for all program elements. PSM program elements and other facility programs and procedures important to quality assurance, will be specifically cross-referenced; and, the cross-reference will be maintained by the Quality Component for future audit. 1 The program elements will be conducted in accordance with approved, written procedures. Training to these procedures will be conducted to ensure the program operates effectively. The program will require documented records to demonstrate compliance Lwith program requirements. The program will include a. level of checks and balances through functional separation and audit. The program will be developed to incorporate quality-at-the-source concepts. Routine quality assurance for safety systems may be performed by the functions responsible for operating the systems. i The. program will embrace issues identification, remedial actions, and management control elements to ensure that deficiencies, deviations, and defective equipment and components are disclosed, and corrected, in a timely manner. [ The' program will be forward-fittmg upon implementation. It will be a bounding { assumption that existing systems were appropriately designed, installed, and operated in l accordance.with applicable requirements and acceptable practices. Existing systems will not be back-fitted except for component replacement, system modification, and/or actions arising from internal investigations and/or external disclosures such as NRC Information Notices. Such back-fitting will always be at the discretion of the Plant Manager, as advised by the Engineering and Regulatory Components. Docket No. 70-1151-Initial Submittal Date: 30APR90 Page No. 3.7 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

Until a system's Integrated Safety Assessment (ISA) is completed, safety-significant controls identified through enhancements of the system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or new controls identified through - configuration management reviews of modifications of the system, will be verified, as necessary, to assure they match the requirements identified in the design criteria. That is, all such controls will be examined in the "as built" condition, and pre-operationally tested, to validate the design and to verify the quality of the installation and the reliability. of the controls. Other specified safety-related controls for a system will be scheduled for "as built" examination and pre-operational testing at the discretion of the cognizant Regulatory Engineer Function. 3.4 ' PROCEDURES, TRAINING AND QUALIFICATION At the Columbia Fuel Fabrication Facility (CFFF), procedures, training and qualification ' j will be integrated into a combined process to assure that environmental protection, health, l integrated safety, quality, and safeguards programs are being conducted in accordance l with Westinghouse policies, and in accordance with commitments to Regulatory s Agencies. Elements of this integrated process will be developed by knowledgeable d Component staff, will be reviewed and approved by cognizant individuals in affected Components, and will be authorized for implementation by Component Management at a 1 level that is responsible and accountable for the operations covered. 3.4.1 PROCEDURES Operations to assure safe, compliant activities involving nuclear material will be conducted in accordance with approved procedures. Approved procedures will be maintained and controlled by an Electronic Procedure System. Approved procedures will provide the basis for training of all personnel involved in operations with nuclear material at the facility. Structured hazards analyses, as conducted in accordance with Chapter 4.0 of this License Application, will include human factors analysis of applicable procedures, as described in Section 3.5 of this License Application. (a) Regulatory-Significant Procedure Structure CFFF procedures will be classified into three general categories: l j (a.1) - Category-1 Procedures Category-1 procedures will be for use by the Regulatory Component. The salient utility of such procedures will be to provide health, integrated safety, and l t p Docket No. 70-1151 Initial Submittal Date: ,30APR90 Page No. 3.8 l -License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

n t safeguards training and instructions for Regulatory Functions. They will be prepared, and approved for issuiag, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and, will be reviewed, and approved ll for issuing, by the cognizant Regulatory Component Manager. The Category-1 scope will group sets of procedure.= into such subcategories as: Administration; Health Physics: e l e. Nuclear Criticality Safety Environmental Protection . Safeguards [ Shipment and Transportation; Instruments; l-Surveys; e Dosimetry; Bioassay; and, e Laboratory Practices Changes to Category-1 Procedurm will be prepared, and approved fcr issuing, by Regulatoy Functions assigned by a cognizant Regulatory Component Manager; l and will be reviewed, and approved for issuing, by the cognizant Regulatory L Component Manager. I i (a.2) Category-2 Procedures l i Category-2 procedures will be for use by individuals outside the Regulatory - Component, and deal exclusively with regulatory practices. The salient utilities of such procedures will be to provide health, integrated safety, and safeguards l training and instructions for Engineering, hm'ufacturing, and Quality Functions; i y and, for use by these Functions in preparing Category-3 Procedures. They will present regulatory guidance methodology acceptable to the Regulatory I Component.- They. will be prepared, and approved for issuing, by Regulatory l Functions assigned by a cognizant Regulatory Component Manager; and, will be reviewed, and approved fc-issuing, by the cognizant ReFulatory Component l Manager. The Category-2 scope will be similar to, and may in many cases overlap, that for Category as applicable to use outside the Regulatory Component. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.9 l u l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l r

Changes to Category-2 Procedures will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and, will be myiewed, and approved for issuing, by the cognizant Regulatory Component Manager. (a.3) Category-3 Procedures Category-3 procedures will be for use by responsible individuals outside the Regulatory Component. The salient utility of such procedures will be to provide training and instructions - including health, integrated safety, and safeguards - l for the Operations, Maintenance, Inspection, and Analytical Services Functions. They will be prepared, and approved for issuing, by Component-Functions assigned by a cognizant Component Manager, based ' on consideration of applicable Category-2 Procedures and/or consultation with cognizant Regulatory Component Engineers; and, will be reviewed, and approved for issuing, by the cognizant Component Manager. The scope of Category-3 Procedures will be as determined by the cogmzant ] Component Manager. Changes to Category-3 Procedures will be prepared, and approved for issuing, by Component Functions assigned by a cognizant Component Manager, and will be reviewed, and approved for issuing, by the cognizant Component Manager. (b) Issuance, Approval, and Communication of Contents of Procedures Acceptable practices for environmental protection, health, integrated safety and l safeguards activities will be provided to operations Components in documented . procedures that are approved, by the Regulatory Component, for electronic issue. Contents of these procedures will be communicated to operations personnel, by Component Management, through incorporation into specified operating and/or L quality assurance procedums.

Regulatory-significant practices in operations and quality assurance procedures, and changes to such procedures, will be issued by cognizant Components in i

i ' accordance with documented policies for procedure preparation, review, and approval. Specifically, Regulatory Component approvals will be required for all I regulatory aspects-of procedures, and. their changes, involving the storage, handling, processing, inspection, and/or transport of nuclear materials.- l Component Management will be responsible for assuring and documenting that contents of these procedures are communicated to appropriate personnel through I Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.10 l l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

- training programs, access to the Electronic Systems, and/or posting of instructions. (c) Procedure Review Frequencies Maximum frequencies of reviews-for-updating for regulatory-significant procedures will be: Annual, for Category-1 and Category-2 Procedures; and, Biennial for Category-3 Procedures. e -(d) Procedure Compliance A formal system will be maintained to enable employees to report imdaquate procedures, and/or inability to follow procedures, to their First Level Managers for follow-up action. First Level Managers will enable, and require, compliance with all regulatory-significant procedures. This will be accomplished by providing ready employee access to procedures, requiring documented employee procedure review and acknowledgement, then evaluating employee performance with respect to procedure compliance on a continuing basis. Employees will receive additional instruction, if determined necessary by the First Level Manager evaluations; and, if procedures are deliberately or repeatedly violated, disciplinary action will be taken in accordance with established Westinghouse policies. 3.4.2 TRAINING AND QUALIFICATION ' Training will be provided for every individual in the Columbia Fuel Fabrication Facility (CFFF), commensurate with their duties. Formal training programs will be developed and implemented to enhance and augment procedure review and. acknowledgement described in Paragraph 3.4.1(d) of this Chapter, and training responsibilities described in Chapter 2.0 of this License Application. Such training programs will be performance-based; and as such, will incorporate the structured elements of job and task analysis, learning objectives, instructional methodology, implementation, and evaluation and feedback. In addition, training of Nuclear Criticality Safety Function Engineers will include qualification by cognizant Regulatory Component Management that goes beyond the' position requirements described in Chapter 2.0 of this License Application. The programs will be structured ruch that specified training and qualification requirements will be met pt:or to safety-significant positions being fully assumed, or covered tasks being independently performed. Training records will be maintained in accordance with Section 3.8 of this Chapter. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.11 l - License No. SNM-1107 Revision Submittal Date:,23JUL98 Revision No. 14.0 l l L1

.(a)' General, Topical, and Refre. sher Training All new employees will receive training relative to safety aspects concerning radiation and radioactive materials; risks involved in receiving low level radiation exposure; basic criteria and practices for radiation protection, nuclear criticality safety (based upon selected guidance from ANSI /ANS-8.20-1991, facility operating experience, and area specific requirements), chemical and fire safety, maintaining radiation exposures and radioactivity in effluents. As Low As - Reasonably Achievable (ALARA), and material safeguards.. Facility visitors will either be provided with equivalent training (commensurate with their visit's scope); and/or, will be escorted by trained employees. Employees or visitors for whom respiratory protection devices might be required, within the scope of their work, will receive pre-work training in the proper use of such devices. - Employees designated to take part in emergency response to facility accidents 'or incidents will receive training commensurate with their assigned activities during such response. P Employees who work with nuclear materials will receive regulatory refresher training on a biennial basis. This training will consist of: Providing each employee with a current revision of the Regulatory Affairs -e Training Manual; Presenting each employee supplementary videotaped instruction on general regulatory issues; and, Requiring each employ:e to successfully pass an examination. e The Training Manual will include such subjects as: ALARA; General health physics practices; e Health physics rules and recommendations; e Area-specific health physics practices; t e l General nuclear criticality safety practices; Area-specific nuclear criticality safety practices; Industrial safety and hygiene, and fire safety, practices; e Chemical Area work practices; e Radiation risks; Docket No.' 70-1151 Initial Submittal Date: 30APR90 Page No. 3.12 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l e

4 e Emergency planning; and, Safeguards. . Employees who are absent from_ the facility during scheduled regulatory refresher training will receive such training within one month of their return to work. 1 (b) Training and Qualification'of Nuclear Criticality Safety Function Engineers 1 Nuclear Criticality Safety Function Engineers will develop skills and abilities directed by the cognizant Regulatory Component Manager, who will evaluate fundamental development methodologies for applicability and utilization on a case-by cases basis. Examples of development methods include: I 'e ' A nuclear criticality safety short course; Westinghouse auditing certification; e American Nuclear Society Standards development and review; Facility criticality safety handbook development and review; e A stmetured hazards analysis coune; - e A structured human factors course; and, e Criticality safety calculations certification. e Demonstrated performance of Nuclear Criticality Safety Function Engineers skills . and abilities will be formally reviewed and documented by the cognizant Regulatory Component Manager and the senior Regulatory Component Manager. . Performance evaluated by the Managers, for review on a case-by-case basis, will E include: Reports of internal audits and inspections conducted; Feedback from worker training presented; e Criticality safety analyses and evaluations perforned. e i Qualification of each Nuclear Criticality Safety. Function Engineer will be formally documented by the cognizant Regulatory Component Manager and the senior Regulatory Component Manager - prior to the Function position being fully assumed, or crucial tasks being independently performed. (c) Training and Qualification of Health Physics Technicians Training and qualification prerequisites for a Health Physics Technician will p include, as a minimum, a high school diploma or equivalent. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.13 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l L

Health Physics Technicians will develop sicitis and abilities, as directed by the cognizant Regulatory Component Manager. Methods evaluated by the cognizant Manager for qualification, on a case-by-case basis, will include: Documented acknowledgement of applicable procedures; Emergency preparedness training; and/or e- ' Applicable skills competency training. ' 3.5 - HUMAN FACTORS Human factors concepts will' be employed at the Columbia Fuel Fabrication Facility (CFFF), in recognition of how the total job environment - areas, equipment, training, and procedures - shapes the expectations, thoughts, and decisions of employees who work wil: !icensed materials. A human factors awareness will be developed at various levels of the organization, and structured human factors analyses will be' performed. Because the operating philosophy of the organization is strongly embodied in procedures, .as described in Subsection 3.4.1 of this Chapter, procedures will receive particular human factors attention. E 3.5.1. DEVELOPMENT OF HUMAN FACTORS AWARENESS . To enable integration of human factors concepts into facility operations, an initial, formal L course - prepared and presented by recognized human factors experts - will be provided for the Plant Manager; all Engineering, Manufacturing, Regulatory, and Quality Component Managers; and, designated Functions from these Components. The course l will address the following elements, including exercises to enhance learned skills: (a) Process Safety Management; - (b) Human Factors Concepts; E r (c) Performance Shaping Factors For Hardware; (d) Performance Shaping Factors For Procedures; l (e) ~ Analysis Preparation; (f) Error-Likely Situations;. I l (g) Procedure Analysis Techniques; (h) Worker Self-Checking Techniques; and, Docket No. 70-1151 Initial Submittal Date: __30APR90 Page No. 3.14 l 1 - License No. SNM-1107 - Revision Submittal Date: 23JUL98 Revision No. 14.0 l

t (i) Supervisor Coaching Principles. I ' 3.5.2 STRUCTURED HUMAN FACTORS ANALYSIS A part of the CFFF Integrated Safety Assessments, described in Chapter 4.0 of this License Application, will include a' structured human factors analysis of assessed system procedures. These analyses will be led by an individual who has completed a formal human factors course. The analyses will embrace the following: (a) Using Procedure-Specific Guide Words For Stmetured Analysis Of Procedures. (b) Minimizing Opportunities For Human Errors Of Omission and Commission Related To Procedures. Results of the structured analyses, including findings and recommendations for improvements, will be documented in formal reports to cognizant Component Management. 3.6 AUDITS AND SELF-ASSESSMENTS The bases of the Columbia Fuel Fabrication Facility (CFFF) Audits and Self-Assessment program will be the performance-based reporting process described in Section 3.7 of this Chapter, the performance-based internal inspection and audit program, and facility management self-assessment of regulatory program performance. 3.6.1 PERFORMANCE-BASEDINTERNALINSPECTIONS AND AUDITS -(a) INFORMALINSPECTIONS l Regulatory Component peronnel on duty, including Regulatory Component management, will conduct continuing informal inspections of regulatory program performance in the course of their routine duties. Observed process upsets and procedural inadequacies will be promptly reported to the cognizant First Level Component Manager for remedial action. Repeated upsets and inadequacies will be reported to the cognizant Regulatory Component Manager, who in turn will l report them to increasingly higher levels of Component Management until effective remedial action has been taken. Such repeated upsets and inadequacies will be documented in monthly formal audits to assure applicable tracking and resolutions. I (b) FORMAL AUDITS-i Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.15 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l i

d Cognizant Regulatory Function Engineers will conduct monthly formal audits of regulatory program perfonnance in accordance with a written procedure. The l. auditors will have the technical capability, and will be formally directed.by Regulatory Component management, to find process upsets and procedural inadequacies well beyond those surfaced by simple paperwork reviews. That is, the audits will include reviews of items entered into the performance-based reporting process, and repeated upsets and inadequacies reported to Regulatory Component management, for the areas being audited; and, detailed area walkdowns. Disclosed upsets and inadequacies will be formally documented in a report to cogmzant First level Component Managers; and, will be tracked by the audit team leader until appropriately addressed. 3.6.2 FACILITY MANAGEMENT SELF-ASSESSMENT The purpose of the self-assessment program will be to provide a means to assure that deficiencies in regulatory performance are identified and corrected to Westinghouse management standards. The Plant Manager will document CFFF policy on the' purpose and objectives of self-assessment to Component Managers, including aggressive demand for quality assessment performance. The management self-assessment. organization will be the Regulatory Compliance Committee (RCC) described in Chapter 2.0 of this License Application. RCC members will be provided with the Nuclear Regulatory Commission Staff's views concerning self-assessment - particularly, that the function of such assessment will be to aggressively disclose and. forcefully report identified process upsets and ' procedural inadequacies before they self-reveal and/or Regulatory Agencies find them. L On a semi-annual basis the following assessment parameters will be summarized and trended by the Regulatory Component: A summary of items documented in the performance-based reporting process; I A summary of upsets and inadequacies documented in performance-based internal audit reports; Facility Collective Dose Equivalent;. Facility average Total Effective Dose Equivalent; Top 10 facility workers' Total Effective Dose Equivalents; Overexposure; A - Regulatory Agency notifications; ] Ratio of Recordable Incident Rate to SIC code average; Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.16 l ' License No.- SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l I

s Imst time accidents per production hour; Results of Special Nuclear Material Physical Inventory (annual); e Emergency response team activations; e L Radioactive emissions in gaseous effluents; f. Radioactive emissions in liquid effluents; l~ Radioactive material transportation incidents; and, e l Regulatory Agency violations. e i The summaries and trends will be formally reviewed by the RCC, particularly for need to be addressed by initiatives of the Safety. Margin Improvement Program described in Chapter 3.0 of this License Application. i i 3.7 INCIDENT INVESTIGATIONS At the Columbia Fuel Fabrication Facility (CFFF), the organizational stmeture described in Chapter 2.0 of this License Application, and procedures in accordance with Subsection 3.4 of this Chapter, will provide for: systematic investigation of abnormal events; making L decisions on corrective measures to prevent recurrence of such events; and, follow-up on the implementation of the preventive measures. Further, the CFFF will have in-place a . structured methodology for determining and categorizing the root cause(s) of the i failure (s) that led to investigated events. ) t 3.7.1 INTERNAL REPORTING OFINCIDENTS L A formal system will be maintained to enable employees to report process upsets and procedure inadequacies to their First I.evel Managers for ' follow-up action; and, ) employees will.be instmeted in its use. Documentation of this performance-based. reporting process will provide for the following information: Event identification number, date, and time. Names of the report originator and the First Level Manager, shift number, and event description; Immediate action taken by the First level Manager; e Explanation of ultimate event closure; and, - Acknowledgement of closure (and date acknowledged) by the cognizant Engineering Function Engineer, the cognizant Regulatory Function Engineer, the . originator's First level Manager, and the originator. 1 l Potential safety-significant reports will be forwarded to the Regulatory Component for l evaluation and determination of necessity for action by the incident review committee, as ' described in Subsection 3.7.2 of this Chapter. All documentation of the performance-i Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.17 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l !~ex__

based reporting process for an area will be reviewed as a part of the formal audits of the area, as described in Paragraph 3.6.1(b) of this Chapter. 3.7.2 STRUCTURED INCIDENT EVALUATION 8 An incident review committee -- comprised of the Engineering Component Senior Manager,. the Manufacturing Component Senior Manager, and the Regulatory Component Senior Manager - will determine if reported process upsets and/or procedure inadequacies are to undergo structured incident evaluation. Structured incident evaluations will be maintained by a datapack process. Documentation of this process will provide for the following information: Results of a Root Cause Analysis, led by an individual with formal training in conducting such an analysis, including recommendations; Status of corrective action (s) implementation; Regulatory assessment; e Notification documentation; ( e Training documentation; e Plant-wide applicability assessment; and, Miscellaneous information pertaining to the incident and/or the evaluation. 3.7.3 ' NOTIFICATION OF REGULATORY AGENCIES Cognizant Regulatory Agencies will be promptly notified of major safety incidents in accordance with all requirements from 10 CFR Parts 20 and 70.~ In particular, as points of additional clarification,- the NRC Operations Center will be notified and/or the NRC Region II Office will be apprised of the following types of incidents, within the time limits prescribed (Note: The incident time limit " clock" starts when a qualified Emergency. Coordinator classifies an event, or when a cognizant Regulatory Function Engineer arrives on-site and makes the initial " eyes-on". assessment of the safety of the errant condition, whichever comes first.): (a) 1-Hour Notifications (NRC Operations Center) '(a'.1) Any incident for which an Alert or Site Area Emergency lus been declared, as l prescribed by the Site Emergency Plan described in Chapter 9.0 of this License. l Application. i (a.2) Any incident involving Quality Level A systems, for which accident controls i L cannot be initiated, whether or not regulatory limits are exceeded. L (b) 4-Hour Notifications (NRC Operations Center) Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.18 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l 4 -a

'(b.1) 'Any incident involving Quality level B systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded. (b.2) Any nuclear criticality safety incident, in an analyzed system, for which less than previously documented double contingency protection remains (multi-parameter i control or single-parameter control) and: l '(b.2a) Greater than a safe mass is involved and double contingency protection is not restored within four (4) hours; or, i (b.2b) Greater than a safe mass is involved and controls are restored within four (4) hours, but:

i. Only single contingency protection is restored; or, ii. Double contingency protection is restored but multiple original controls under both contingencies. vere modified or replaced.

(b.3) Any determination that a criticality safety analysis or evaluation was deficient,' or that a particular system was not previously analyzed; and, that less than two unlikely, independent, and concurrent changes in process conditions would be required before a criticality accident would be possible. (b.4) Any unanticipated /unanalyzed nuclear criticality safety incident, or incident-involving a previously.unanalyzed system, for which the severity and remedy are not readily determined. (c) 24-Hour Formal Notifications (NRC Operations Center). (c.1) Any incident for which the work area is unavailable for normal use for 24-hours, following a loss of radioactivity contamination control, where the unavailability is due to the contamination. (c.2) - Any incident for which Quality Level A or B system safety equipment is not performing its intended function. .(c.3) Any incident for which an employee, having significant removable radioactivity - contamination (see definition, Section 1.4), receives medical treatment outside of facility contamination control areas (c.4) Any incident for which a fire or explosion damages nuclear fuel and its processing equipment or container is breached. L - Docket No. 1151 Initial Submittal Date: 30APR90 Page No. 3.19 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l l

9 (c.5) Any nuclear criticality safety incident, in an analyzed system, for which less than previously documented double contingency protection remains (multi-parameter control or single-parameter control) and: (c.5a) Less than a safe mass (see definition, Section 1.4) is involved; or i (c.5b) Greater than a safe mass is involved, but a sufficient number of the controls that l were lost are restored within four (4) hours such that double contingency protection is restored. (d) Next-Working-Day Informal Apprisal (NRC Region II O' ffice) ) i (d.1) Any determination that a particular system was not previously analyzed; even l though, at least two unlikely, independent, and concurrent' changes in process conditions would be required before a criticality accident would be possible. (d.2) Any unanticipated nuclear criticality safety incident, or incident involving a previously unanalyzed system, for which the severity _ and remedy are readily . determined. (e) A procedure _will be prepared, maintained, and followed - in accordance with - Subsection 3.4.1 of this Chapter - that details the information to be included in a notification. -In particular, each notification of a nuclear criticality safety incident will include '(where applicable) the following information: Whether the notification is the result of an event, or. of a deficient nuclear. e criticality safety analysis (including the time period for.which the deficiency existed); The safety significance of the incident; Potential criticality-pathways involved, -including brief. scenario (s) of how accidental criticality could occur; Controlled parameters - mass, moderation, geometry, concentration, etc. - e. involved; . Estimated amount, enrichment, and' form of licensed material involved - l e [ including applicable process limits and the percent of worst-case critical mass of the material, in the configuration, involved; EA description of the involved failures or deficiencies - including applicable n e L nuclear criticality safety controls or control systems; and, Corrective actions to restore safety systems, and when each was, or will be, e

impemented.

Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.20 l License No. SNM-1107 ~ Revision Submittal Date: _23JUL98 Revision No. 14.0 l l L

3.8 RECORDKEEPING AND REPORTING - The Columbia Fuel Fabrication Facility will identify, maintain, preserve, control, and destroy records -- as defined in the records management section of the controller's manual - in accordance with the guidelines, procedures, and practices set forth by the Westinghouse Electric Corporation. Such records, specifically required by applicable regulations, will be maintained in accordance with those regulations. Reporting of ( records data will be as prescribed by applicabb regulations. 3.8.1 RECORDS Written procedures, prepared and maintained in accordance with Subsection 3.4.1 of this Chapter, will specify the management program for licensed activity records; including: .~' '(a). Enviromnental Surveys; (b) Radiation And Contamination Surveys; 'l (c) Personnel Exposures; l I (d) Instmment Calibration Results; (e) Nuclear Criticality Safety Evaluations, Analyses and Methodology Validations; ' udit And Inspection Reports; (f) A (g) ALARA Reports;. (h) Regulatory Compliance Committee Meeting Minutes; (1). - Employee Training And Re-training Documentation; l (j)' Records Of Plant Alterations Or' Additions; ) (k) Documentation Of Abnormal Or Atypical Occurrences And Events Associated i. With Radioactivity Releases; 1 (1) Decontamination And Decommissioning Files; and, -(m) Other Such Records Required By the Regulations. These procedures will include Records Flow Schedules, which list: i Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.21 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l L_

l Record category, Name of record; Form numbers; e Retention period in active files; e Retention period in the central records bureau; and, { e Retention period in the remrds center, j Records of tests, measurements, and surveys required to document compliance with conditions.of operating licenses and permits will be retained for at least three years, unless otherwise specified in the regulations. i 1 Records of nuclear criticality safety analyses will be retained-for the lifetime of the {

facility.

t 3.8.2 RECORDS RETRIEVAL ) 1 All retained records will be stored, and maintained readily accessible, in order to meet time restraints relative to their use. Retained records will be as complete and detailed as { necessary to enable traceability to original source data. 1 The records retention system will include the capability to retrieve records within 24-hours for records generated within the past 12-months; and, inside 7-calendar-days for older generation periods. 3.8.3 RECORDS RE-CREATION Prudent measures of protection and redundancy will be afforded such that acts of record alteration or inadvertent destmetion will not foreclose capability for reconstructing a complete and correct set of required records. In cases where protective measures fail, and a particular record is lost or inadvertently i destroyed, a reconstruction may be generated using source data applicable to the time the { subject record was originally created.' When a document is just partially missing, all . salvaged portions will be attached to the reconstruction. -If source data is not available for re creating a missing record, the record may be reconstmeted using inference to data i relative to other documents for similar information and tune periods. f I- .3.8.4 REPORTS I { A' detailed listing of reports required by NRC regulations will be maintained and followed. This listing will document-d Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 3.22 l 3 License No. SNM-1107 - Revision Submittal Date: 23J0ii.9_8_ Revision No. 14.0 l e

Reference to applicable regulations; Descriptions of the reports required; and, e Frequencies at which the reports must be submitted. e . Docket No. 70-1Q11 hiitial Submittal Date: 30APR90 Page No. 3.23 l - License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l c _______________- _ l

CHAPTER 5.0 RADIATION SAFETY I 5.1 ALARA (As Iow As Reasonably Achievable) POLICY The Regulatory Compliance Committee will maintain oversight of CFFF management's commitment of making every reasonable effort to maintain radiation and radioactivity exposures, to employees and the general public, as low as is reasonably achievable (ALARA). 5.1.1 Ultimate ALARA responsibility and authority will be vested-in first level managers, with assistance from upper management and cognizant service groups.~ Specific ALARA recommendations and requirements will be generated by the Regulatory Component, in I the form of environmental protection and radiological safety policies and procedures. ALARA policies, expectations, and goals will be identified, and provided to first level managers, by the Regulatory Compliance Committee. Selected ALARA criteria will be incorporated into operating procedures, as applicable. ALARA ~ considerations will be incorporated-into the : design of new, or. modified, facilities and equipment and will be verified and approved by the Regulatory Component. Individuals responsible for facility and equipment designs will be directed to interface with the Regulatory Component, at the earliest stage of development, to assure that . ALARA philosophies are appropriately incorporated into each project. ALARA principles and requirements will be provided to employees as part of routine training sessions. 1 5.1.2 Short-term. ALARA progress _ will be based on a formal quarterly evaluation and j documentation of ALARA program indicators, by a task group of cognizant Regulatory Component representatives. Such indicators will include, but may not necessarily be - limited to: surface contamination, airborne radioactivity excursions, aiderne radioactivity averages, DAC-hours,L bioassay, Total Effective Dose Equivalent, and effluents; particular attention will be paid to new operations. Results of these evaluations will be documented and, unfavorable trends and/or failure to meet specified goals will be reported as~ a regular agenda item at quarterly Regulatory Compliance Committee meetings. l 5.1.3 Ieng-term ALARA progress will be based on a formal annual report, by the Regulatory L Compliance Committee, to the Plant Manager. This ALARA Report will review Docket No. 70-1151 Initial Submittal Date: _3_0APR90 Page No. __he l License No. SNM-1107 - Revision Submittal Date: 23JUL98 Revision No. 14.0 l f= L --_ _ ___ _ _

t i. I' exposure and emuent data to determine: (1) if there are any upward trends developing in concentrations or quantities of radioactive materials in effluents to the environment, or in personnel exposures for identifiable categories of workers or types of operations; (2) if L such concentrations or quantities in emuents, or exposures, might be -lowered in accordance with the ALARA concept; and (3) if equipment for emuent and exposure l control is being properly used, maintained, and inspected. The ALARA Report will include review of related audits and inspections performed during the reporting period; and will summarize data from the following areas: emuent releases, environmental 4 monitoring, inplant airborne radioactivity, personnel exposures, bioassay results, and unusual occurrences. l . 5.1.4 Implementation of the ALARA program, including commitments in Subsections 5.1.2 and 5.1.3, will satisfy the requirement (10CFR20.1101(c)) for an annual review of the radiation protection program content and implementation. 5.2 RADIATION WORK PERMITS (RWP) 5.2.1 CRITERIA (a) Specific requirements of the Radiation Work Permit (RWP) program will be documented in an approved procedure. I t (b) A Radiation Work Permit will be required for all work for which radiation protection requirements are not covered by operating procedures and one, or . more, of the following conditions is met: (b.1) Release of detectable contamination outside of a Contamination Controlled Area ~ I might result in contamination of personnel or equipment by the work under consideration. L (b.2) The local concentration of radioactive contaminants is predicted to average 50-percent, or more, of Derived Air Concentration (DAC), as a result of the work under consideration. 1 i (b.3) The deep dose equivalent is predicted to exceed 100 millirem in a week. as a result of the work under consideration. L (b.4) The Total Effective Dose Equivalent is predicted to exceed 10-percent of the 10CFR20 limit, as a result of the work under consideration. l f '_ c) . RWP's will be requ".sted by the responsible department, and such requests will be ( submitted to the Regulatory Component for evaluation, preparation and approval. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.1-l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l l l L_.____ __.__..___.__________.____________.____..____.______J

The Regulatory Component will specify applicable protection requirements for the work to be performed. Approvals from the Regulatory Affairs Component and the cognizant first level manager will be obtained prior to starting the activity. -(d) Only personnel who have completed appropriate safety training as detailed in Subsection 3.4.2 of this License Application - will be assigned to perform work under an RWP. (e) A copy of the Radiation Work Permit will be made available to personnel . working under the RWP, and the work will be conducted as specified in the approved permit.

5.3.

VENTILATION SYSTEMS 5.3.1 Ventilation systems will be designed and operated to assure adequate control of process- . generated radioactive dust and particulate matter. Air flows will be typically maintained ) from non-chemical process areas to Chemical Areas. Whenever adverse air flows are detected, corrective actions will be taken as soon as practicable. .j i l 5.3.2 Ventilation systems servicing laboratory-type hoods, production hoods,. and/or other primary enclosures where uncontained nuclear material is handled, will provide a minimum face velocity of 100-linear-feet per minute at all openings during work operations. Face velocity measurements will be made, and documented, on at least a quarterly basis when such equipment is in operation; and, systems found not to meet the q minimum flow velocity will be measured on at least a weekly basis until a documented evaluation demonstrates the minimum flow velocity can be maintained. 5.3.3 When containment of uranium dust by conventional ventilation hoods is not possible, or is impractical, gloveboxes may be used. Ventilation systems for gloveboxes, and similar enclosed devices, will be designed and operated with a nominal negative internal pressure of at least 0.1-inches of water with respect to room air. Gloveboxes will be equipped - . with instrumentation for tueasuring differential pressure, and such instrumentation will be checked for proper operation on at least a monthly basis when the enclosed equipment is in operation. L l When positive-pressure atmosphere control is required for product quality or other approved purpose; and, where there is un-encapsulated radioactive material in positive-pressure gloveboxes, the following criteria shall apply: (a) Gloveboxes will be designed for high integrity confinement. g (b) Gloveboxes will be operated with a maxunum positive internal pressure of 0.1 L i Docket No. 70-1151 Initial Submittal Date: __30APR90 Page No. 5.2 l L License No. SNM-1107' Revision Submittal Date: 23JUL98 Revision No. 14.0 l h. 1 E-1--_ =-_ __ __.-- J

i inches of water with respect to room air. (c). Internal atmospheres will be continuously re-circulated through HEPA filters, at a flow rate designed to attain at least 20-atmosphere-changes per hour. (.d) Alarms will be provided, to indicate when pressure exceeds the pre-set positive L pressure limit. f i . (e) An interlock, or other pressure relief device, will be provided to exhaust the glovebox with a sufficient factor of safety to ensure its continuing integrity. l 5.3.4 - Ventilation hoods and gloveboxes will be constructed primarily of metal, using glass l and/or fire resistant plastic for viewing areas. Plastics will conform to a Class-I fire l~ rating. l 5.3.5 Ducts' will be designed to minimize accumulations of nuclear material, and will be inspected on a routine basis commensurate with the potential for material accumulations. 5.3.6 Exhausts from gloveboxes, hoods, local exhaust enclosures, and similar devices, when employed for radiation protection purposes, will be passed through HEPA filtration. The HEPA-filters will be replaced, either on a routine schedule, or when airborne activity concentrations, hood velocities, differential pressure drops, or particulate penetration measurements indicate that replacement is necessary. The maximum differential pressure permitted across a HEPA filter will be 8-inches of water for negative pressure systems and 4-inches of water for positive pressure systems. 5.3.7 Exhausts from in-plant, recirculating process-air cleaning systems, including gloveboxes, hoods, local exhaust enclosures, and similar devices, will either have their HEPA filters penetration tested, or will be sampled for airborne radioactivity concentrations, on at least a quarterly basis; and, maintenance will be performed on systems found to exceed 25% DAC. 5.3.8 Ventilation control facilities, equipment, and systems, as necessary to minimize . exposures to radioactive materials will be developed and utilized. 5.3.9 Gloveboxes, ventilation hoods, or other containment devices will be installed and used whenever-they are determined to be necessary as a result of radiation protection measurements or evaluations. 5.3.10 The effectiveness of the final HEPA filters, in process ventilation equipment and containment systems, will be determined by in-situ testing, using particulate penetration methods, or other testing means, selected by the Radbition Safety Function. Such testing will be g L Dc:ket No. 70-1151 Initial Submittal Date: J0APR90 Page No. 5.3 l License No. SNM-1107 Revision Submittal Date: 23JUL98 kevision No. 14.0 l

performed following each filter change. 3 5.3.11 Adequacy of ventilation and containment controls within the licensed activity will be

determined by continuous air sampling. The action levels in Subparagraph 5.4.1(i) will be used

' as guidance to determine adequacy of ventilation and coni.inment. 5.4 ' AIR SAMPLING 5.4.1 WORK AREA A1R SAMPLING ('a): All areas wly exposed radioactive materials are handled will be sampled for airborne radio.ctive particulate matter using a combination of fixed location general area air samplers, fixed location breathing zone representative air samplers, or portable air samplers. The type of sampling employed, and location of samplers, will be determined by the Radiation Safety Function. (b) Fixed location air samplers used for the purposes of assessing and assigning operator intake will be located in or around the breathing zone of operator work stations where uranium handling operations are performed, or where short term operations occur frequently. Breathing zone representativeness of these samplers will be estr.51ished when the samplers are initially installed. Samplers will be re-examined for cepresentativeness annually, or whenever substantive equipment or process changes are made (in accordance with Section 3 of Regulatory Guide 8.25, " Air Sampling in the Workplace.") Representativeness studies will be performed in accorciance with methods and acceptance criteria described in Table 2 of Regulatory Guiie 8.25. (c) Fixed location air samplers will be located where potential airborne contamination hazards exist; such that, deterioration in ventilation controls, containment i controls, or operating procedures resulting in significant increases in airborne radioactivity concentrations will be detected so. corrective actions can be instituted. (d) All new operations or substantive modifications to existing equipment will be evaluated by the Regulatory Component, or sampled using portable samplers to assess the need for fixed location sampling stations. (e) Lapel samplers may be used on a limited basis to supplemen: ".her air sampler measurements where work stations are not defined, or for special studies. i .(f) Continuous, alarming air monitors may be used on a qualitative basis to provide early warning for operators in the event of a significant airborne release. Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.4 l p - License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l I

(g) Work-area air samples will be changed out at least once each working shift and allowed time for natural activity decay before processing during production operations, unless a documented evaluation by the Radiation Safety Function demonstrates that another schedule is justified. Samples will be analyzed using calibrated counting equipment; and, airborne activity concentrations calculations will account for filter collection efficiency, self-attenuation, and counter efficiency. Analyses will include radiological counting of the samples to determine radiological concentrations in the work areas. (h) Air samples suspected of reflecting releases and significantly elevated concentrations will be counted as soon as practicable following sample change out, to determine radioactivity concentration. (i) All work-area sampling programs will provide for investigation, special sampling, and/or increased sampling frequency if the activity concentration outside of containment structures (not directly resulting from a specific known cause) exceeds the following action levels (where the DAC for Class Y uranium is used): A single sample collected for 8 hours'or longer exceeds 250% DAC. The monthly average for a sample location exceeds 100% DAC. 'o Operations or equipment will be shut-down, and immediate corrective action will be taken, at locations where a single air sample exceeds 1000% DAC. 5.4.2 Fixed in-plant air and gaseous effluent sampling systems will be subject to surveillance by the Radiation Safety Function. Such surveillance will assure that flow meters are working and properly adjusted, that the vacuum system is intact, and that filter media has been properly installed. 5.4.3 Air flow measurement devices on the fixed in-plant air sampling system, the gaseous effluent sampling system, and environmental air samplers, will be verified annually for l proper operation, and will be replaced as required. 5.5 CONTAMINATION CONTROL 5.5.1 ' CONTAMINATION SURVEYS Contamination surveys will be performed on a continuing basis, to evaluate the potential spread of radioactive contamination. (a) Contamination surveys w31 be performed with sufficient frequency to assure that Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.5 l - License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

maximum acceptable limits are not exceeded. Maximum acceptable limits, and minimum survey frequencies for iloors and other readily accessible surfaces will be as specified in Figure 5-1. Specific portions of a Contamination Controlled Area may be assigned higher limits and/or frequencies, provided a documented evaluation by the Radiation Safety Function has demonstrated that collective protective measures for the subject area will assure compliance with licensed and regulatory requirements. Examples include areas where contamination does not represent the potential for becoming airborne or being tracked and decontamination is impractical (e.g., under process equipment, hoods, etc.). i FIGURE 5-1 CONTAMINATION SURVEY LIMITS AND FREQUENCIES 1 l AREA TYPE ALPHA ACTIVITY MINIMUM . ON SMEAR

• FREQUENCY Change Rooms, and 50 Weekly Eating / Vending Areas Clean Area 200 Monthly Contamination Controlled Area 5000 Biweekly 1

' Units of Disintegrations-Per-Minute Per 100-Square Centimeters ) (b) All new operations will be subjected'to increased contamination surveillance until l experience has shown that the routine schedule will be adequate to protect health i and safety. (c) If the average contamination level in the Contamination Controlled Area exceeds l the specified action level, decontamination will be required within three l working-shifts; and, immediate decontamination, or area isolation, will be mquired if the average contamination level in the Contamination Controlled Area is greater than five times the specified action level. An area not greater than 10-square-meters will be used to determine such average contamination levels. W Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.s l License No, SNM-1107. Revision Submittal Date: 23JUL98 Revision No. 14.0 l L ._.__m__m-

i a given smear exceeds an action level, additional smears may be taken, as l necessary, to demonstrate that immediate. decontamination is not required. Verification surveys will be performed to verify decontamination. In the Clean Areas, decontamination is required immediately following the discovery of any contamination above the appropriate limit. (d) . An alpha smear measurement technique will be used which is capable of detecting 25-disintegrations-per-minute per sample, at a 90-percent confidence level, when surveying clean areas, change rooms, and eating and vending areas. 5.5.2 ACCESS Access to areas in which radioactive materials are used or stored will be controlled. Convenient change rooms and step-off pads will be provided at access points, to prevent l the spread of contamination from Contamination Controlled Areas. .(a) Personnel will be authorized to enter Contamination Controlled Areas, by virtue - of management approval in accordance with the CFFF Physical Security Plan, .j only after completing required radiation protection training. 1 (b) . Access points to Contamination Controlled Areas will be established through change rooms and. step-off pads. Each such access point will define a contaminated side and an uncontaminated side, with a step-off area provided between the two sides. 1 (c) Each access point to the Contamination Controlled Area will be posted in accordance with 10CFR20.1902, except for 10CFR20.1902(e). In lieu thereof, a sign bearing the legend "Every container or vessel in this area might contain radioactive material" may be posted at entrances to each such area in which radioactive materials are used, or stored. (d) Access to Contamination Controlled Areas, including the Chemical Manufacturing - Area and other areas involved in the processing and storage of unencapsulated . radioactive materials, will require the use of protective clothing. (e) Protective clothing will be provided for personnel entering the Contamination Controlled Area. This will include labcoats, coveralls, shoecovers, safety shoes, and/or other specified garments - consistent with an individual's work assignment. Street clothing, of persons to be dressed completely in protective clothing, will be stored on the uncontaminated side of the change, line. Used - protective clothing will be stored on the contaminated side of the change line until l-collected for laundering. Contamination limits for protective clothing will be L Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.4 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

consistent with the limits in Figure 5-1. (f) Personnel survey instmments will be provided in change rooms and at step-off pads, for use by personnel leaving Contamination Controlled Areas. Instruments will be checked for proper operation daily, during production operations, by the Radiation Safety Function. (g) Instructions will be posted at exit points from Contamination Controlled Areas, which describe survey techniques, procedures for decontamination, and what to do in event of survey instrument malfunction. 5.6 EXTERNAL EXPOSURE 5.6.1 PERSONNEL MONITORING DEVICES Film badges or thermoluminescent dos! meters, provided by a commercial supplier which is NVLAP certified, and capable of detecting and measuring beta-gamma and x-radiation, will be provided to individuals specified by the Radiation Safety Function. These badges or dosimeters will be evaluated at least quarterly, or more frequently as specified by the Radiation Safety Function. In addition, neutron detection capability will be available, for use as specified by the Radiation Safety Function, and will be evaluated at least quarterly. 5.7 INTERNAL EXPOSURE The Regulatory Component will perform a biennial evaluation of vendors used to analyze bioassay samples. Such evaluations will also be performed if substantive program i anomalies are disclosed. The evaluations will consider the need for " spike" and J " replicate sample" submittals. Tle invivo counter will be calibrated at least annually; and, will be functionally tested each day the counter is in operation. A bicassay program will be maintained to evaluate the effectiveness of material control and personnel protection programs, to evaluate intakes exceeding action levels specified in Subsection 5.7.3 and to assess dose used to determine compliance with applicable occupational dose equivalent limits (diagnostic bioassay only). Samples will be analyzed by a qualified laboratory using fluorometric methods for routine urine samples and radiometrically for diagnostic urine and/or fecal samples. The primary method of assessing and calculating intake and Committed Effective Dose 2quivalent (CEDE) will be by the measurement of breathing zone representative air sampling results. 5.7.1 INVITRO BIOASSAY Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.8 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

1 (a) Routine urmalysis samples will be collected for the purpose of tracking and evaluating potential long term accumulation and retention of radioactive material 'in individuals. Samples will be submitted annually by individuals required to be ' monitored for internal radiation exposures. The Radiation Safety Function will perform evaluations of statistically meaningful results and prescribe subsequent 1 actions to be taken based on there evaluations. 1 (b) . Work activity restrictions will be imposed, and diagnostic bioassay samples will be requested, when air sampling indicates exposures exceeding the action levels in Subsection' 5.7.3 may have occurred. Such bioassay measurements, invivo measurements, air sampling measurements or any combination of these measurements will be used to assess intake and dose used to demonstrate compliance 'with the occupational dose equivalent limits in 10CFR20. .(c) Acceptable biokinetic models and intake assessment methods specified in Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," will be used in the interpretation of bioassay measurements. (d) Baseline urinalysis measurements will be performed for individuals required to be monitored for internal exposure prior to initial work activities that involve exposure to radioactive material. Termination measurements will be performed, when practical, if an individual is no longer subject to the bioassay program due to changes in the individual's employment status (such as termination of employment or changes in the individual's assigned duties). 5.7.2 INVIVO BIOASSAY (a) Routine uranium lung burden evaluations will be performed for the purpose of tracking and evaluating potential long term accumulation and retention of radioactive material in individuals. Lung burden evaluations will be performed l annually for individuals required to be monitored for internal radiation exposures. The phosry methoe of evaluating uranium lung burden will be the performance of invivo lung counts. Evaluations of air sample data, operator stay-time, routine invitro (fecal and urine) sampling, or any combination of these methods will be performed to evaluate potential uranium lung burden for individuals not able to be invivo counted (i.e., claustrophobic individuals). The Radiatioa ' Safety Function will perform evaluations of statistically meaningful results and prescribe subsequent actions to be taken based on these evaluations. Docket No. 70-1151 Initial Subinittal Date: 30APR90 Page No. 5.9 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l t l

(b) Diagnostic uranium lung burden analysis will be performed when an individual exceeds the intake or dose action levels specified in Subsection 5.7.3. or the action level for nontransportable uranium exposures specified in Subsection 5.7.3. Such invivo measurements, invitro bioassay measurements, air sampling measurements or any combination of these measurements will be used to assess intake and dose used to demonstrate compliance with the occupational dose equivalent limits in 10CFR20. (c) Baseline lung burden evaluations will be performed for individuals required to l be monitored for internal exposure prior to initial work activities that involve exposure to radioactive material. Termination measurements will be performed, when practical, if an individual is no longer subject to the bioassay program due to changes in the individual's employment status (such as termination of employment or changes in the individual's assigned duties). 5.7.3 RADIATION EXPOSURES (a) Individuals likely to receive greater than 10% of the applicable Annual Limit on Intake (ALI) will be monitored for intakes of radioactive material. Suitable and timely measurements of radioactive material in the air of the work area, measurements of radionuclides in the body, measurements of radionuclides excreted from the body, or any combination of airborne concentration, invivo and invitro bioassay measurements will be used to monitor intakes to individuals. (b) Committed Dose Equivalent (CDE), Committed Effective Dose Equivalent (CEDE), and Total Effective Dose Equivalem (TEDE) occupational doses will be calculated in accordance with 10CFR20 and acceptable methods described in Regulatory Guide 8.34, " Monitoring Criteria and Methods to Calculate Occupational Radiation Doses." (c) Intakes to Class D uranium will be limited to less than 10 milligrams uranium per week per individual. (d) Work restrictions and diagnostic evaluations will be performed when an individual receives a single intake of greater than 40 DAC-Hours exposure to Class W and/or Class Y uranium or 20 DAC-Hours Class D uranium (which corresponds to approximately 5 milligrar.^s Class D uranium @ 3.5 weight % U-235 enrichment). (e) Work activity restrictions will be imposed when an individual exceeds 80% of applicable limits; i.e.,0.8 ALI,1600 DAC-Hours,4.0 REM CEDE for inhalation exposures to Class W and Y uranium,4.0 REM TEDE,4.0 REM DDE,40 REM Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.10 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

CDE,etc.). i 5.8 RESPIRATORY PROTECTION l 1 l A policy statement will be written on respirator usage and will include the following: 5.8.1 Engineering controls and administrative procedures will be provided to minimize the need for respiratory protection. 5.8.2 Respiratory protection equipment will be used in accordance with written procedures, and individuals using respiratory protection will be trained in accordance with the criteria in a 10CFR20, Subpart H. 5.8.3 Only respirators certified by the National Institute for Occupational Safety and Health /Mine Safety and Health Administration (NIOSH/MSHA) will be used. 5.8.4 Protection factors iam Appendix A,10CFR20 will be used when assigning actual intakes. 5.8.5 Personnel authorized to use respiratory protection equipment will be fit-tested biennially. 5.8.6 Personnel authorized to use respiratory protection equipment will be trained in the applicable requirements biennially. 5.8.7 Personnel will be required to test respirators for operability immediately prior to each use. 5.8.8 Written policies and procedures will cover the following: (a) respirator selection, fitting, issuance, maintenance and testing; (b) supervision and training of personnel; (c) monitoring, including air sampling and bioassay; (d) recordkeeping; (e) determination by a physician prior to the hdtial fitting of respirators, and periodically at a frequency determined by a physician, that the individual user is medically fit to use the respiratory protection equipment; l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.11 l . License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l i b

CDE,etc.). 5.8 RESPIRATORY PROTECTION A policy statement will be written on respirator usage and will include the follo'ving: 3.8.1 Engineering contro's and administrative procedures will be provided to minimize the need for respiratory protection. 5.8.2 ' Respiratory protection equipment will be used in accordance with written procedures, and individuals usig respiratory protection will be trained in accordance with the criteria in 10CFR20, Subpart H. 5.8.3 Only rt.nirators certified by the National Institute for Occupat!> al Safety and Health /Mme Safety and Health Administration (NIOSH/MSHA) will be used. 5.8.4 Protection factors from Appendix A,10CFR20 will be used when assigning actual intakes. 5.8.5 Personnel authorized to use respiratory protection equipment will be fit-tested biennially. 5.8.6 Personnel authorized to use respiratory protection equipment will be trained in the applicable requirements biennially. 5.8.7 Personnel will be required to test respirators for operability immediately prior to each use. 5.8.8 Written policies and procedures will cover the following: (a) respirator selection, fitting, issuance, maintenance and testing; (b' supervision arJ training of personnel; (c) monitoring, including air sampling and bioassay; (d)' recordkeeping; (e) determination by a physician prior.to the initial fitting of respirators, and periodically at a frequency determined by a physician, that the individual user is medically fit to use the respiratory protection equipment; i Docket No. 70-1151, Initial Submittal Date: 30APR90 Page No. 5.11 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No.14.0 l

4 (f) use of process or other engineering controls, instead of respirators; (g) routine, nonrottine and emer;;ency use of respirators; and (h) periods of respirator use and relief from respirator use. 5.9 INSTRUMENTATION 5.9.1 RADIATION PROTECTION INSTRUMENTS Instmments used for radiation protection measurements will have capabilities as follows; however, more than one instmment may be utilized to cover the specified range: (a) Portable Survey Instruments - Alpha,100 to 1.0E06 disintegrations pcr minute; . Beta-Gamma, 0.1-millirem per hour to 300-REM per hour; neutron, 0.5 to 5 MREM per hour. (b) ' laboratory Assay Instruments - Alpha,10-percent of the regulatory limit for Derived Air Concentrations (DAC), for sampling periods of 8-hours or more. Radiation protection instruments' will be calibrated on a routine schedule established by the Radiation Safety Function. The i,chedule will require calibration following initial instmment acquisition; and. thereafter, at minimum, following major repairs, semiannually, or the ' manufacturer's recommendation, whichever is lesser. Aipha counting instruments used in the Radiation Safety Function Laboratory will be checked each working day, when in use, to determine background activity; and, a calibrated source will be counted to assure proper instmment functioning. A voltage ' plateau, defining the proper counting voltage for each such laboratory alpha counting instrument, will be determined quarterly. Instrument calibration records will be maintained for a period of at least three years. Operability of portable sur<ey instruments will be detetmined Mer to each use. 5.10 SUMMING INTERNAL AND EXTERNAL EXPOSURES 5.10.1 RADIATION DOSES .(a) internal and external occupational radiation doses will be combined in accordance with criteria in 10CFR20 and applicable guidance contained in Regulatory Guide 8.34, " Monitoring Criteria and Methods to Calculate Occupational Radiation Doses," and Regulatory Guide 8.7, " Instructions for Recording and Reporting l Occupational Radiation Exposure Data." L Docket No. 1151 Initial Submittal Date: 30APR90 Page No. 5.12 l License No. SNM-1107 Revision Submittal Date: 23JUL98_ Revision No. 14.0 l c

5.10.2 DOSE TO EMBRYO / FETUS (a) Radiation dose to the embryo / fetus will be calculated in accordance with applicable guidance in Regulatory Guide F.36, " Radiation Dose to the Embryo / Fetus." n ) 4 l 1 l I l l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 5.13 l - License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

.9 CHAPTER 8.0 . FIRE 3AFETY At the Columbia Fuel' Fabrication Facility (CFFF), fire protection will be achieved by combinations of fire protation measures and systems. Such measures and systems will be ' designed and maintained in accordance with industry standards and ' prudent practices. The' standards and practices most often consulted will be those of the National Fire Protection ~ Association (NFPA). 8.1 STRUCTURE OF THE FIRE SAFETY PROGRAM A' multi-component, Engineering and Regulatory, team will be empowered by facility management to formally evaluate and defne the CFFF Fire Safety Program. At least one team member will have documented certification as a Process Hazards Analysis Izader. The team will provide the results of the evaluation and defined program structure to Engineering and Regulatory Component management. The results and structufe will also be retained for future inspection by Regulatory Agencies. As dermed by the fire safety program evaluation team, the program structure will be maintained as follows: 8.1.1 FIRE PROTECTION PROGRAM (a) The charter of the Safety Committee described in Section 2.2 of this License Application will specify its responsibilities to include the fire protection program. The Committee will meet at a minimum frequency of once per calendar quarter. l The Committee's fmdings, conclusions, and recommendations will be reported directly to the Plant Manager following each meeting. Day-to-day ~ protection reviews will be the. responsibility of the cognizant ' Regulatory Engineer Function. (b) The initial fire hazard analysis will be as documented in the Westinghouse Nuclear Fuel Columbia Site Evaluation Report (March,1975). A supplement to this analysis will be as documented.by Impell Corporation in the Fire Hazard Analysis for Westinghouse Commercial Nuclear Fuel Division Columbia Plant (June,1987). : The current fire hazard analyses will be as found in the Pre-Fire Plans for the various areas of the facility. The Pre-Fire Plans will be updated as described in Paragraph 8.1.9 (c) of this Chspter. l-Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.0 l l L i % No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

. 9; ~ .4 . (c) A fire protection preventive maintenance program will be in place; and, relevant documentation will be maintained on the maintenance system described is Section 3.2 of this License Application. (d) Review and control of modifications of the facility or processes to minimize fire hazard will be as described in Section 3.1 of this License ' Application. (c) An approved cutting and welding procedure, welder training programs, and hot work permits will be provided to control welding and torch cutting activities. Non-routine use of flammable materials will - be controlled by the same i precautions utilized for routine use of such materials. Flammable liquids will be retained in containers and/or cabinets designed for such purpose; arxi, additional { precautions will be as specified by the cognizant Regulatory Engineer Function. (f) Inspection, testing, and maint:runen of fire protection equipment will be covered by the fire protection preventive maintenance program. (g) Basic -fire protection training will be covered in new-hire. and contractor orientation programs The emergency response tear will be given documented trairdng, and drills.will be conducted, u part of the Emergency Management f Program described in Chapter 9.0 of this License Application. j (h)

A system will be provided to enable reporting of fire incidents to First Level 1

Management, as described in Section 3.7 of this License Application. Fire alarm pull stations will tv strategically located throughout the-facility. Areas with potential fire hazanis will be equippi with appropriate fire detection and/or suppression systems. The Security Function will be responsible for announcing alarms and alerting-personnel to iire incidents. Following announcement of an alarm, instructions .will be given to apprise personnel of any necessary actions. Approved procedures, as described in Section 3.4 of this License Application, will define reporting guidelines and investigation requirements for fire incidents. j l (i) An emergency exercise, in accordance with the E -.gf Plan described in Chapter 9.0 of this License Application, that includes facility evacuation, will be conducted on'a biennial basis. At' times prescribed by the cognizant Regulatory Engineer Function, a fire scenario will be included. i . (j) - Approved procedures, as described in Section 3.4 of this License Application, i [ I o i Docket No. 70-1151 ' Initial Submittal Date: 30APR90 Page No. 8.1 l 'I h ' License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

[. L.,..- will prescribe the housekeeping practices for the facility. Good housekeeping techniques will be practiced at the facility as an integral part of the Westinghouse total quality culture. j '8.1.2 ADMINISTRATIVE CONTROLS (a) Program management will follow the organizational. responsibilities and authorities structure described in Section 2.1 of this License Application. (b) Plant audits and inspections of fire protection will be performed at two levels, as follows: (b.1) Formal monthly audits, supplemented by informal inspections, as described in Section 3.6 of this License Application, will include fire safety. Safety observers will also perform formal and informal inspections, of specified process areas, that include fire protection, combustible loading, and housekeeping status. (b.2) American Nuclear Insurers (ANI) performs independent fue protection, prevention, and brigade inspections of the facility. Action plans will be developed to address findings arising from such inspections. 8.1.3 BUILDING CONSTRUCTION -(a) The construction standards for the facility manufacturing areas were those in place when the aren were originally constructed. The building structural members were built using non-combustible, or limited combustible materials. When the building structure is me * ;ed or expanded, prevailing NFPA code requirements will be met. (b) Fire response areas will subdivide specified processes and materials involving fire hazards, to confine fire to its area of origin and prevent its spread. In particular, the following building units will be such areas: Solvent extraction; - Boiler rooms; Incinerator; Warehousing areas (MRO Storeroom, Product Storeroom); L Process control rooms; Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.2 l L 1 License No.- SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l 1 ! ~. __________________w

-D ' ] i Switchsear rooms; . Computer room; l Maintenance shops; e' Fire pump areas; f + Office areas; l Emergency generator rooms; and, d Compressor rooms. ~) These areas will conform to: location and manning requirements; fire barrier ratings; fire detection requirements; sprinkler (and other fire suppression method) -{ specifications; combustible material inventory controls; container and containment j specifications; wiring grades; and/or, housekeeping practices - as defurd and ) documented. .(c) . To minimize exposure fire risk, the facility will employ guidance from the NFPA 30 and NFPA 80A standards. (d) To enable rapid personnel egress from the f~:ility in event of a fire, the fwility will employ guidance from the NFPA 101 standard. (e) Smoke vents will be located in the' mechanical manufacturing areas. 1(f) Hidden (concealed) spaces will be routinely addressed as part of the monthly l formal audits and informal inspections, described in Section.3.6 of this License Application.- -(g)- Lightning protection of steel buildings will be maintained with grounding straps; and, specified equipment will be grounded. _j (h) Drainage provisions will be addressed by properly sized floor drains; sumps will be used where specified. ~(i)~ Electrical installations and wiring will conform to ' dustry standards. m 8.1.4 ' VENTILATION SYSTEM 1 Docket No. 70-1151 . Initial Submittal Datei 30APR90 Page No. 8.3 l i License No.1 SNM-1107 ' Revision Submittal Date: 23JUL98 Revision No. 14.0 l

O (a) Facility ventilation systems will be designed for fire protection. -(b) Class 1 UL-586 (or equivalent) final HEPA filters will be used. (c) Barrier penetrations will employ fire dampers designed to specifications. (d) At:tomatic closing will be required for fire doors and dampers. (e) Space heating furnaces will be built to industry and NFPA 70 standards. 8.1.5' PROCESS FIRE SAFETY (a) Prunary chemicals used at the facility will be evaluated for their fire hazards, and their control will be defined. In particular, the following chemicals will be controlled:

• Nitric acid;
• Sulfuric acid;
• Anhydrous Ammonia; and,
• - Hydrogen.

Use of such chemicals will conform to: operator hazard recognition; training in . safe handling and spill prevention techniques; storage; containment; maintenance; leak testing; and/or, safety shut-off valve verifications -' as defined and documented. (b) Processes involving use of flammable liquids will not be introduced to the facility j mtil they are evaluated, and their control will be defined and documented. j Processes involving use of flammable gases will be evaluated, and their control l will be defined - in particular, the following controls will be applied to . flammable gas processes, as defined and documented: I' Sintering furnaces, as specified, will be upgraded to meet NFPA 86C standards. Combustible gas analysis will be performed prior to hot (open flame) work, as specified on work permits. Docket No,' 70-1151 Initial Submittal Date: 30APR90 Page No. 8.4 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l

' Sintering furnaces will be provided with flame curtains designed to continually burn-off excess hydrogen gas upon release. from furnace atmosphere. Process interlocks will be eniployed to assure proper operation of the flame curtains. (c)' The fire hazard in handling of uranium oxides will be evaluated. Non-combustible materials will be specified for powder handling systems where the potential for spontaneous exothermic reachon needs to be considered. Where - high density polypropylene containers are used for storage and transport of active - uranium oxides, operators will be trained to recognize hazardous powder characteristics, and will be instructed to monitor for exothermic reactions in such containers. !(d) Machining operations on combustible metals at the facility will be evaluated for their fire hazards, and their control will be dermed. In particular, the following operations involving potential for zirconium metal fmes will'be controlled by approved procedures, as described in Section 3.4 of this License Application: Fuel rod repair stations; Final fuel assembly loaders; Laser welders; Zirconium grid strap production areas; Mechanical development laboratories; and, e' i e Tool rooms. Such areas will conform to: containment, ventilation, and filtration; and/or, fire extinguisher requirements - as dermed and documented. - (e) The facility incinerator will be isolated by a rated fire barrier as dermed and ' documented. Incinerator exhaust will be ducted to a filtration and sampling system prior to release to the environment. Incinerator' exhaust will be passed through a cooling water media for cooling and dust separation. (f)- Boilers and boiler-furnaces will be evaluated, and their control will be defined - . in particular, the following cordrols will be applied, as defined and documented: - Boilers will be physically separated from the process buildings; and, thus, L, l I Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.5 l ) Ik~me No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l L_ _= __ - -_-

i l ) may be contained in non-fire-rated boiler houses. Construction and operation of boiler-furnaces will be'in accordance with indwtry standards. Fuel storage tanks will be separr.ted from the boiler house; and, fuel lines will be marked for. identification, and located to minimize damage potential? i (g) Stationary combustion engines will be evaluated, and their control will be defined -- in -'particular, the following. controls will be = applied, as definal and documented: i Emergency gemrator areas will be protected by a sprinkler fire suppression system. l Stationary combustion engines will be located in rooms constructed of non-combustible materials. Engme rooms will be configured such that process-generated dusts or 4 flammable vapors are not present. i l<uel storage tanks will be located' outside engine rooms; and, will be - e constmeted in'accordance with industry standards. Engine exhaust systems will be designed to prevent ignition of combustible material - by contact with hot metal surfaces, or by leaking' exhaust gases and sparks. Engine rooms will be ventilated to minimize accumulation of combustible vapors. Such ventilation systems will be automatically activated when engines are started. (h) Storage and handling of flammable and combustible liquids and gases will be evaluated,' and their control will be defined - in particular, the' following controls will be applied to fhmmable and combustible liquids, as defined and documented: Combustible liquid storage systems will be designed and reaintained as specified. Indoor storage of flammable and combustible liqaids will be as specified; and, fire extinguishers will be available. p I Docket No. 70-1151'

Initial Submittal Date:

30APR90 Page No. 8.6 l l License No. SNM-1107 ' Revision Submittal Date: 23JUL98 Revision No. 14.0 l { N

O. -.9 j Supports of aboveground storage tanks will be protected from potential

exposure to fires.

1 Aboveground storage tanks will be provided with emergency relief vents, in accordance with industry standards. Construction, installation, operation, and maintenance of bulk gas storage, loading, and dispensing systems will be in accordance with prudent - industry practice. (i) Glove boxes will be evaluated for fire hazards, and controls will be defined - in particular, the following controls will be applied, as dermed and documented: Glove boxes will be constructed of nons:ombustible material, except for gloves and windows, as specified. Explosive mixtures in glove boxes will be prevented, and inert gas or dry 1 air atmospheres will be employed, as specified. (j) Fim protection methods for laboratories handling radioactive materials will be in accordance with industry standards, as dermed and documented. 8.1.6 FIRE DETECTION AND ALARM SYSTEMS (a) ' Automatic fire detectors will be installed in specified areas with substantial: combustible loadings, and-with infrequent occupancy. (Unless such areas are covered by automatic fire suppression systems). (b)- Automatic flammable vapor and gas detectors may not be installed for hydrogen ' systems that _have been evaluated and for which it has been determined and ' documented that potential for leakage is minimal and sufficient dilution air is- . pmsent to prevent formation of explosive ndxtures. -(c) ~ Audible fire alarms will be installed in specifk:! locations throughout the facility; - visual' alarms will be installed in high noise areas, as specified. Such alarms will be monitored by a continuously manned, central control station that indicates fire detection system and zone status. (d). Manual fire alarm actuators (pull-boxes) will be installed in specified locations throughout the facility.- Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.7 l License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l n

h'E 8.1.7 < FIRE SUPPRESSION EQUIPMENT (a) Selection of equipment for suppression of fire will take into account the severity

of the hazard; the type of activity to be performed, the potential consequences of a fire, and the potential consequences of use of the suppression equipment

~(particularly, risk of an accidental criticality). ,(b) . Automatic sprinkler systems will be selected and designed in accordance with industry standards, as defined and documented. Automatic sprinkler systems will be specifically excluded from areas where moderation is a principal nuclear . criticality safety concern, and/or in areas with a concentration of energized electrical equipment. .(c) Standpipe and hose systems will be selected and designed in accordance with industry standards, as defined and documented. Standpipe and hose systems will have read.ily accessible hose outlet locations. (d) Portable fire extinguishers, of sufficient capacity and type of suppression agent used, will be available and maintained throughout the facility. Portable fire extinguishers will be selected and deployed in accordance with industry standards, . as defined and documented. . 8.1.8 FIRE PROTECTION WATER SYSTEM (a)- Water supply for fire protection systems e e: ;c r sured as described in Subsection 8.2 of this Chapter. (b) Fire pump installation will be adequate to deliver water to hydrants, standpipes, and sprinkler systems -- a described in Subsection 8.2 of this Chapter. (c) Alternative power for fire pumps will be provided. Diesel pumps wi!! be test-started on a weekly basis; and, two sets of batteries will be provided for back-up starting. Emergency response personnel will be trained to start the pumps ' manually. l(d) The water distribution system will be designed such that failure of a single component will not forclose ability to deliver fire suppression water to the facility. 8.1.9 FIRE HAZARD ANALYSIS ' (a).. Fire hazard rnalyses for the facility will be part of the pre-fire plans addressed in l Section 8.1.10 of this Chapter, as defined and documented. Integrated safety. - Docket No.. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.8 l - License No. SNM-1107 Revision Submittal Date: 23JUL98 Revision No. 14.0 l t Ii

f't assessments, described in Chapter 4.0 of this License Application, will include - fire hazard analyses. 4 i (b) Fire hazard analyses will be reviewed by cognizant Regulatory Component i Management. In particular, disclosed. deficiencies will be reviewed and considered, by cognizant Component Management, for inclusion in the safety 1 margin improvement program described in Chapter 3.0 of this License Application. (c) Fire hazard analyses will be updated when the cognizant Regulatory Engineer Function determines that a significant number of minor facility changes have accumulated to warrant updating; and, whenever there has been a significant change of the facility, or its processes or inventories. ' 8.1.10 PRE-FIRE PLAN (a) The facility will maintain, on file for insr:,etion by. Regulatory Agencies and ready to use, pre-fire plans that provice information needed by fire-fighting personnel responding to an e,mergency. (b) Contents of Pre-Fire Plans The facility will be divided into logical planning areas. I.a:ations of logical plan'iing areas, response team assembly points, response team assembly po!m coverage areas, and fire hydrants will be .. n indicated on site sketches,. l Basic response team responslbilities will be assigned. and team checklists will be provided. ' Locations of fire detection anxi protection devices will be listed. l 1 Each area plan will detail: a d:scription of the area; expected occupancy; potential le;,ations. ibt trapped occupants; any disabled personnel that might require arsistance in an einergency; utilities information, construction information, managem h. 'nd phone numbers; } plan update scheduling; hazardous materials basic utformation;' strategy l consider;;tions; and, additional comments deemed relevant by the cognizant Regulatory Engineer Function. (c)' . Pre-fire plans (and revisions to such plans) will be prepared by the cognizant l i . Docket No. 70-1151 Initial Submittal Date: 30APPS_0 Page No. 8.9 l l' License No. - SNM-1107 Revision Submittal Date: 23JU198 Revision No. 14.0 l j

~ ' Regulatory Engineer Function, and copies will be provided to the off-site fire l department most likely to respond to a call for assistance. 8.1.11 FIRE EMERGENCY RESPONSE TEAM (a) Organization and equipment of the emergency team will be maintained to respond to facility fires, as part of the Emergency Management Program addressed by Chapter 9.0 of this License Application. (b) Training will be provided to the emergency team, to enable performance of duties in response to facility fires, as part of the Emergency Management Program addressed by Chapter 9.0 of this License Application. 8.2 FIRE SUPPRESSION SERVICES The 10 inch water main that supplies process and drinking water to the site will also i supply two water tanks, with a combined capacity of 450,000 gallons available for use in ) fire fighting. The tanks are checked weekly, and topped-off with water as required. l (Based upon historical data, a minimum water volume of 85-percent of tank capacity will i thus be' maintained.) The associated fire pump is rated at 1,000 gpm flow at 100 psi pressure. Multiple 6-inch fire hydrants, with 2.5-inch hose connectors, will be located at strategic locations about the facility site; and, multiple 1.5-inch standpipes will be strategically located throughout the facility. 1 l f a 1 l 1 1 l Docket No. 70-1151 Initial Submittal Date: 30APR90 Page No. 8.10 l f ( License No. SNM-1107 Revision Submittal Date: 23JUL98, Revision No. 14.0 l -}}