ML20211A098

Rev 16.0 of Page V & Pages from Chapters 1.0,2.0,3.0,10.0 & 11.0 of Application for Renewal of SNM-1107,correcting Typo in Possession Limits & Incorporating NRC Inspector Suggestions for Clarification of Employee Training

ML20211A098
Person / Time
Site:	Westinghouse
Issue date:	08/16/1999
From:	Robert Williams WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
CON-NRC-99-035 NUDOCS 9908230185
Download: ML20211A098 (63)

Text

\\

k Westinghouse Commercial 21 clear Raw R Sectric Corporation Fuel Division ffaNYeNo'"*"**"

NRC-99-035 August 16,1999 Director Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

CHANGED PAGES; LICENSE NUMBER SNM-1107; DOCKET 70-1151 Westinghouse Electric Company hereby submits (six copies of) a proposed Revision 16.0 of page y and pages from Chapters 1.0, 2.0, 3.0,10.0 and 11.0 of the Application for Renewal of a Special Nuclear Materials License for the Commercial Nuclear Fuel Division operations at the Columbia, South Carolina Fuel Fabrication Facility. The substance of ik changes correct a typographical error in possession limits, and incorporates NRC Inspector suggestions A,r clarification of employee training and liquid waste treatment commitments. The changes also update the organization structure, and incorporate other minor administrative revisions.

If you have any questions, please contact me at (803) 776-2610, Extension 3393.

Sincerely, WESTINGHOUSE ELECTRIC COMPANY 2='

Robert A. Williams l

- Licensing Coordinator Docket 70-1151 l

License SNM-Il07 cc:

U. S. Nuclear Regulatory Commission ATTN: Mr. Harry Felsher rO Licensing Section 1, Licensing Branch s

FCS&S Division, NMSS

- 11545 Rockville Pike Mail Stop T8D14 ON (pO, n Rockville, MD 20852-2738 Enclosures l

9908230185 990816 PDR ADOCK 07001151 C

PDR

p 39 l-i TABLE OF CONTENTS L

PAGE NUMBER AND TITLE TABLE OF CONTENTS......................................................................................... i REVISION RECORD............................................................................................. iii CHAFTER 1.0 -

GENERAL INFORMATION.................................................... 1.0 1.1 FACILITY AND PROCESS DESCRIPTION................................

1.0 L

1.2 INSTITUTIONAL INFORMATION.......................................... 1.4 1.3 SITE DESCRIFFION............................................................. 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 SAFETY COMMITTEES........................................................ 2. 8 CHAFTER 3.0 CONDUCT OF OPERATIONS................................................. 3.0 3.1

~ CONFIGURATION MANAGEMENT........................................ 3.0 3.2 MAINTENANCE.................................................................. 3.2 3.3 -

QUALITY ASSURANCE........................................................ 3 3.4' PROCEDURES, TRAINING AND QUALIFICATION................... 3.8 3.5 H UMAN FACTORS............................................................. 3.14 3.6 AUDITS AND SELF-ASSESSMENTS...................................... 3.15 i

3.7 INCIDENT INVESTIGATIONS.............................................. 3.17 3.8-RECORDKEEPING AND REPORTING.................................... 3.20 J

CHAPTER 4.0 INTEGRATED SAFETY ASSESSMENT...................................... 4.0 CHAFTER 5.0 RADIATION S AFETY........................................................... 5.0 1

5.1 ALARA (As Iow As Reasonably Achievable) POLICY................. 5.0 5.2:

RADIATION WORK PERMITS (RWP)...................................... 5.1 5.3 VENTILATION SYSTEM S.................................................... 5.2 5.4 -

' AIR SAMPLING................................................................... 5.4 5.5 CONTAMINATION CONTROL............................................... 5.5 i

5.6 -

EXTERNAL EXPOS URE....................................................... 5. 8 5.7 INTERNAL EXPO SURE........................................................ 5.8 j

5.8 ~

RESPIRATORY PROTECTION.............................................. 5.1 1 5.9 INSTRUMENTATION.......................................................... 5.12 5.10 SUMMING INTERNAL AND EXTERNAL EXPOSURES............ 5.12 I

- Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No. _i License No! SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No. 16.0 F

t I

l i

i

L

.f TABLE OF CONTENTS (Cont'd)

L l

NUMBER AND TITLE PAGE CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY......................................... 6.0 6.1 PROGRAM ADMINISTRATION.............................................. 6.0 6.2.

CONTROL METHODOLOGY AND PRINCIPLES....................... 6.2 6.3 ALARM SYSTEM................................................................ 6.19 6.4 CONTROL DOCUMENTS..................................................... 6.20 CHAFTER 7.0 CH EMICAL SAFETY............................................................ 7.0

. 7.1 CH EMICAL SAFETY PROGRAM............................................ 7.0 7.2 CHEMICAL SAFETY HAZARD EVALUATIONS....................... 7.0 7.3 CHEMICAL SAFETY PROGRAM STRUCTURE........................ 7.1 J

7.4 ADDITIONAL CHEMICAL SAFETY COMMITMENTS............... 7.2 I

CHAPTER 8.0 FIRE S AFETY...................................................................... 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 i

. 9.1 EM ERGENCY PLAN........................................................... 9.0 9.2 EMERGENCY EQUIPM ENT.................................................. 9.0 CHAPTER 10.0 ENVIRONM ENTAL PROTECTION........................................ 10.0 10.1 EFFLUENT AIR TREATM ENT.............................................. 10.0 1

10.2 LIQUID WASTE TREATMENT FACILITIFS............................10.0 10.3 SOLID WASTE DISPOSAL FACILFFIES..................................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 CONCEPTUAL DECOMMISSIONING PLAN........................... 1 1.0 11.2 DECOMMISSIONING FUNDING FLAN AND FINANCIAL ASSURANCE MECHANISM................................................ 11.1 CHAPTER 12.0 AUTHORIZATION S AND EXEMFrlONS................................ 12.0 12.1 AUTHORIZATIONS............................................................. 12.0 12.2 EXEM PTION S.................................................................... 12.5 p

f Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.,,ji License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

s r

v

n.

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 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%

iii,1.12,3.18, and 3.19 Incorporation of a definition, (Reprinted all document and incident notification pages in Microsoft Word criteria, recently approved format) by NRC Staff.

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 authorization, and expand another authorization to enable cement manufacturing with CaF2.

8.0 11AUG97 -

iii,2.4 and 8.1 (Reprinted Change emergency exercise all document pages in frequencies for consistency Microsoft Word format.)

with Emergency Plan.

- Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

iii License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

E 9.0 23SEP97 iv, Chapter 6 To respond to NRC Staff request for additional information.

To revise table to correlate to CSE orgaruzation and clarify discussions regarding margin of safety with respect to normal operations, expected I

process upsets and credible process upsets.

10.0 31 MAR 98 Table of Contents i-iv To replace Revisions Chapter 6.0, Numbers 2.0, 4.0, 6.0, and (NOT YET APPROVED.

9.0; and respond to SMIP BEING REVIEWED BY NRC STAFF) initiative regarding SNM-1107, Chapter 6.0. (Chapter l

l 3.0 shown with bars & 6.0 l

Major Rewrite).

11.0 03APR98 Table of Contents i-iv, To reflect common 3.18 - 3.20.

understanding on notification.

12.0 30JUN98 Table of Contents (iv),

To update and enhance Chapter 1.0 (1.12),

Integrated Safety Assessment l

Chapter 4.0 (all).

commitments.

13.0 13JUL98 Table of Contents (iv).

To update and enhance License Safety Condition S-2.

l 14.0 23JUL98 Table of Contents (iv).

Chapter 1.0 (1.4,1.5);

To 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.

i-Chapter 3.0 (3.3).

To delete pellet carts from programmed maintenance.

Chapter 3.0 (3.15).

To clarify commitment to formal audits.

)

Chapter 5.0 (5.9, 5.10).

To expand commitment for invivo bioassay.

Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

iv License No. SNM-1107

- Revision Submittal Date: 16AUG99 Revision No. 16.0

f.

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.

15.0 12FEB99 Table of Contents (v)

To clarify material Chapter 1.0 (1.4) possession limits.

'16.0 16AUG99 Table of Contents (v)

To correct typographical error Chapter 1.0 (1.4,1.5,1.9, and clarify employee training 1.10, 1.12).

and liquid waste treatment l

Chapter 2.0 (2.0, 2.1) commitments.

Chapter 3.0, (3.11, 3.12)

To update organization Chapter 10.0 (10.0, 10.1) structure and incorporate Chapter 11.0 (11.2) admimstrative revisions.

l l

[

i i

I l

l l

l

\\

l l

l Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No. _v License No. SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No. 16.0 i

1 CHAlrrER 1.0 GENERALINFORMATION 1.1

' FACILITY AND PROCESS DESCRIPTION The Columbia Fuel Fabrication Facility (CFFF) of the Commercial Nuclear Fuel 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, uranium hexafluoride; converting the hexafluoride to produce uranium ' dioxide powder; 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 technically sound radiation and environmental protection, nuclear criticality safety, industrial safety and health, SNM safeguards, and quality assurance controls described in detail in this License Application.

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 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 EMERGENCY PLAN described in Chapter 9.0 of this License Application.

- 1.1.1-SITE UTILITIES AND SERVICES (a)

Electrical Supply The CFFF will be served by a single, 115,000 volt, electrical supply line. Four diesel-powered standby generators will be installed and maintained to meet the emergency electrical power requirements of the site in the event of a temporary l

L outage of the normal supply source. Emergency power will be automatically.

provided to crucial process equipment; emergency lighting systems; cooling system pumps; all fim 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

License No. SNM-Il07 Revision Submittal Date: 16AUG99 Revision No. 16.0

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 contammation, from process areas will be routed through HEPA filtration, to remove entrained uranium particulates, prior to discharge to the environment. Exhausts containing uranium in soluble form will be passed through aqueous scrubbers, precedmg the HEPA filters.

Following filtration, the gases will be continuously sampled, to enable analyses for assuring compliance with the limits specifkxl 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 consist of filtration, flocculation, lime addition, distillation, and precipitation (in a 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 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 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, 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 U-235, and stored to await incineration. Noncombustible wastes, and selected Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.1 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

i t

combustible wastes, will be packaged in compatible containers, compacted when appropriate, measured to verify the uranium content, and placed in storage to await shipment for further treatment, recovery, or disposal.

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 that the above referenced controls are effective.) Wastes designated for disposal will be packaged l

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 within the manufacturing or waste storage buildings; or, they may be stored outdoors, if protected from the elements.

l l

Wastes consigned to disposal will be shipped to a licensed burial facility.

l Shipments will be made in compliance with all applicable NRC, DOT and State l

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 l

FOR THE COLUMBIA FUEL FABRICATION FACILITY, dated April 1,1987, J

l 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 Low Strategic Significance, in accordance with applicable requirements of 10CFR73 and 74.

1.1.2 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 as uranium hexafluoride, uranium nitrates, uranium oxides; and/or contained in pellets, fuel rods, fuel assemblies, samples, scrap, and wastes; (2) Receipt, handling, and Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.2 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0 l

p 1

L 1

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 inspecdon, and final fuel assembly; (5) Quality I

assurance and control inspection activities; (6) Analytical Services Laboratory operations - including wet-chemistry and spectrographic techniques; (7)

Metallurgical Laboratory operations - including sample preparation, polishing, testing, and examination; (8) Chemical Process Development operations -

I meluding laboratory-scale process research, prototype development, and equipment I

including check-out; (9)' Mechanical Process Development operations laboratory-scale research and development; (10) Health Physics Laboratory operations - including sample preparation and analysis, instrument repair and calibration, respirator fit-testing, and bioassay sample and sealed-source storage; l

(11) In-house, and contracted, scrap recovery operations - including scrap batch processing, solvent extraction, coated-pellet recovery, scrap blending, and l.

hydrofluoric acid recovery; (12) UF6 cylinder washing, hydrostatic testing and re-certification; (13) Equipment and facility maintenance activities; (14) Equipment and facility decontamination activities - including clothing; (15) Waste storage and disposal preparation operations - including HEPA filter testing, conversion liquid waste treatment, advanced waste-water treatment, lagoon storage, incineration, radioactive waste packaging for disposal, and calcium fluoride disposition; (16) l Ancillary mechanNI operations - including non-radioactive component fabrication and asserably; sad (17; Shipping container and overpack refurbishment.

(a.2) The licensed activity may hiso perform work for other Westinghouse Divisions, or outside customers, which is within the authorized capabilities of the facility.

l 1

(b)

Material Possession Limits and Constraints The following will be the maximum quantities of Special Nuclear Material that may j

be possessed by the licensed activity at any one time; and, constraints for I

procurement, use, and transfer of such material.

(b.1) Material possession limits - (1) 5-grams of U-233 in any chemical or physical form, limited to laboratory use as individual 1-gram maximum quantities 'm ventilated hoods; (2) 350-grams of U-235, as uranium of any enrichment, in any chemical or physical form; (3) 75,000-kilograms of U-235, as uranium enriched l

' Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

1.3 i

License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

[

I to no greater than 5.0 weight-percent, in any chemical or physical form except metal; (4) 1.5-grams of Pu-238/239 as sealed sources; and, (5) Transuranics and fission products in feedstock, not to exceed 3,300 Bq Alpha / KgU or 440,000 Mev Bq Gamma / KgU (i.e., the limits on alpha and gamma activity specified for

" enriched reprocessed UF6" in ASTM C996-% " Standard Specification for Uranium Hexafluoride Enriched to less Than 5% 2"U") - not to exceed a total mass of 5-grams of Plutonium.

3 (b.2) Material constraints - The procurement of Special Nuclear Materials will be in accordance with licensed activity needs. Production, utilization, and/or significant loss of special nuclear materials 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 (CFFF). Westinghouse Electric Company LLC is controlled and owned by BNFL Nuclear Services Inc. (BNSI), a wholly owned United States subsidiary of British Nuclear Fuels plc (BNFL). In accordance with the requirements of 10 CFR 70.22(a)(1), the following additional information is submitted:

1.2.1 - APPLICANT AND STATE OF INCORPORATION Westinghouse Electric Company LLC

' Delaware 1.2.2 LOCATION OF THE PRINCIPAL OFFICE Monroeville, Pennsylvania 1.2.3 NAMES (CITIZENSHIP) AND ADDRESSES OF PRINCIPAL OFFICERS

. Charles W. Pryor (USA)

President and Chief Exe.utive Officer Westingham Electric Company

' Westinghouse Energy Center P. O. Box 355

Pittsburgh, Pennsylvania 152304)355 1

l l

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.4

. License No. ~ SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

)

i i

{L,

L James'A. Fici (USA)

General Manager, Commercial Nuclear Fuel Division

- Westinghouse Columbia Plant l

P. O. Drawer R l:

Columbia, South Carolina 29250 Jack B. Allen (USA)

CFFF Plant Manager-

- Westinghouse Columbia Plant l

Drawer R-Columbia, South Carolina 29250 l

1.2.4 CORPORATE CONTACT FOR LICENSING MA'ITERS GriffHolmes Manager, Environmental Health and Safety

- Westinghouse Energy Center P. O.- Box 355 Pittsburgh, Pennsylvania 15230-0355 l.2.5 SITE CONTACT FOR LICENSING MA'ITERS Robert A. Williams Licensing Project Manager

' Westinghouse Columbia Plant Drawer R Columbia, South Carolina 29250 li6 ADDITIONALINFORMATION Additional corporate financial and business information is provided in the Westinghouse Annual Report, available from:

Westinghouse Electric Company P. O. Box 355 Pittsburgh, Pennsylvania 15230-0355 1.3 SITE DESCRIPTION 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 southeast of the city limits of Columbia (see Figures 1.1 and 1.2) along South Carolina Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

1.5 i

License No. SNM-1107' Revision Submittal Date: 16AUG99 Revision No.16.0

)

~

FIGURE 1.1 CFFF SURROUNDING AREA N

W- -E SOUTH CAROLINA m,,, ? ?

'?

?

?--

One inch equals approalmately 35.6 miles g" ', c," JarCW

]c,is "' /,'

i E

,,1 3

l 2)

Chet g

3,c,g eanargart 4

p settwee '

, s7 i

l 48

+-

ee.n c,,,,

"'""'1 Av 23 1

es i

i 2.-

(,3 Cia

• • =9 '

si a c' 1

o somme 4

/

Y

. 3 m"anea.

"'* W

{

i,"'t n

g m

n g-t

~e w

. 33 5

,,s.

nN mi '

. _,y

"'""'",, y

1.

b LOW t,,,.s.oi,.5

,re, mgy..

s2 i

,- 3

• \\,

. 4

.,.u.-

u.

23 5'

OLUMBIA as u

n is

\\A so 3"

I

~

18 F

g,g,g,g m

suas a ern.im,M City Aye6 eld m

8 women atos n,

.,* w.la se u

n 8i M

,.y'8"'8i sp l

14 26

• ' I t***

5 ansu poeth metiN. \\.

k El 5

s5W

',,,'E y/

u

's g,a awr*

s 22 m,

- 4.,e' )&,s g

..c siW g,,, '".

g. is

,s

.i o

'l

,\\ e %me.... p,,M' is 1

y um

,gg,,)

w-.

i a.*e

,,,a.,

- a,,

,i

<=* ag u./

g q....

r... J

= b a"O., J C/.L....

.., a e m

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.6 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

1 FIGURE 1.2 CFFF PROPERTY BOUNDARY OIf,o"c'au,.ra. s.c.

ele

{-

v I

s

~~'\\

s

~~]

/

r'

• • j

.,s I

\\

J l

I I

/

I 2

\\

• e l

]

]

%g

+

l\\

l s&

A CHF-92002 SHT-1

• See Figure 1.3 Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

1.7 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0 i

s Highway 48. The region around the site is sparsely settled, and the land is characterized by timbered tracts and swampy areas, penetrated by ummproved roads.

Farms, single-family dwellings, and light commercial activities are located chiefly along nearby highways.

The site is bordered by abutting properties, as presented in the PHYSICAL SECURITY PLAN described in Paragraph 1.1.l(e) 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 landscaped areas. A site plan is shown in Figure 1.3.

Details of the CFFF location, including 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 defmed and used as indicated:

ALTERNATIVE ACTIONS - Tests, procedures or other practices that may be substituted for prescribed activities as deemed appropriate by the Regulatory Component.

In such case, a detailed analysis will be performed and documented by the cogmzant Regulatory Functions. This analysis will include a comparison of the proposed action with that specified in the license; and, a demonstration that action levels 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 contammation on floors and accessible surfaces is high, and protective clothing is required; such as, the UF6 Bay, the Conversion Area, the Pelleting Area, the Rod leading 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 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.

I Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.8 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

e FIGURE 1.3 SITE PLAN tm uvn l@ l D

[-, O n --~

.@e f41.,

l ** =

• I staa

/5 '#

g;-, s

/t

'O O

p'

,, f0 M

\\n's" e 0 88

'2"""

r<

a g

u O

/(%

)

ou Lotus n,sa 5

O

{S A u

i,-

f, LEGEND:

• !?.KJ.>i

(,

<=.

@ EiF G' iib -

c.,-,,..,..._.,

I Docket No.

70-1151 Initial Subnitta) Date:

30APR90 Page No.

1.9 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

{

l 1

SITE PLAN KEY Si:

Shipping / Receiving point for UF6 Cylinders.

R1 j

S2 :

Shipping / Receiving point for Uranyl Nitrate.

R2 S3:

Shipping / Receiving point for U Powders, U Pellets, R3 :

U Scrap, and U Waste.

54 :

Shipping / Receiving point for U Rods and U Assemblies.

I R4 SS :

Shipping / Receiving point for miscellaneous U Samples R$

and U Standards.

1.

Fuel Manufacturing Building.'

i 2:

UFe Storage Pad.'

3:

Gatehouses (4).

l 4:

Administration Building.

5:

Parking Area.

6:

Lagoons (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.'

' Nuclear Material Routinely Contained l

l Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

1.10 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0 l

l

F L

i CONTAMINATION CONTROLLED AREA - An alternate name for the Chemical Area.

CONTROLLED ACCESS AREA - A physically defined area, represented on three sides oy a seven-foot hip barrier of Number-11 American Wire Gauge fabric-fence topped by three strands et barbed wire, and on the fourth side by the Admuustration and Mam Manufacturing Building. This area is the " Controlled Access Area" described in the l

Physical Security Plan.

~ 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 true enrichment level is 5.0 w/o U-235 or less can not be rejected at the 0.05 level of significance.

EQUIVALENT EXPERIENCE - When used in a personnel qualification context to equate experience with education, eight years of applicable experience is equivalent to a 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 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 days; MONTHLY means twelve per year, with each covering a span of 40-days or less; QUARTERLY means four per year, with each covering a span of 115-days or less: SEMIANNUAL means two per year, with each covering a span of 225-days or less; ANNUAL means once per year, 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 covermg 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 responsibility, and authority to make and implement a component, having the capability,igned duties; such as the Environmental Prot decisions required to carry out ass Function, Radiation Safety Function, Nuclear Criticality Safety Function, Chemical Safety Function, Fire Safety Function, and Safeguards Function of the Regulatory Component.

LICENSED ACTIVITY - That combination of personnel, plant, and equipment established by Westinghouse Electric Corporation to carry out the processing of radioactive material authorized by this License Application.

MAY - Denotes implied permission by NRC Licensing Staff to take a stated action or course.

PORTABLE AIR SAMPLE - An air sample that is not integrated into the plant's central air sample vacuum system.

l Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

1.11 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 l

i l

RADIATION WORKER-Any individual who, in the course of employment, is likely to receive an annual occupational dose in excess of 100-millirems.

REGULATORY-SIGNIFICANT PROCEDURES -- Those procedures that contain, in whole or in part, actions that are important to environmental protection, health, safety, l

and/or safeguards.

{

i RESTRICTED AREA - Areas such as the Manufacturing Building, or equivalent areas, to which access is restricted by physical or administrative methods and which is monitored on a scheduled basis by the site Security Function.

j SAFETY MARGIN IMPROVEMENT CONTROLS - Controls that provide cost effective enhancements to the safe and effective operation of a process. These are controls that enhance an existing and adequate margin 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 UO2) and reflection.

Optimum moderation and material density are assumed.

SAFETY-RELATED - Relevant 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 safe s,

chemical safety, and fire safety. These controls, which include both Safety-cant" and " Safety Margin Improvement" controls as sub-sets, will be identified ugh an integrated safety analysis which documents the design safety basis for a panicular process.

SAFETY-SIGNIFICANT - Relevant to systems crucial or important to safety (e.g., in the context of quality assurance).

SAFETY-SIGNIFICANT CONTROLS - Controls cmcial or important to, or deemed desirable for, the safe and effective operation of a process, and an adequate safety margin for the process. An adequate safety margin is made up of those controls necessary for the safe operation of the process plus those controls identified to ensure regulatory compliance.

UNRESTRICTED AREA - An area, access to which is neither limited nor controlled.

WILL - Denotes a mandatory requirement to take a stated action or course.

l i

Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

1.12 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

CHAirrER 2.0 MANAGEMENT ORGANIZATION i

2.I ORGANIZATIONAL RESPONSIBILITIES AND AUTHORITIES The Westinehm Electric Company is divided into divisions. One such division is the Commercial Nuclear. Fuel Division (CNFD), which encompasses commercial activities 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 Westinghouse Electric Company, 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 i

directly to the President and Chief Executive Officer of Westinghouse Electric Company.

Within CNFD, the primary 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, quality, and 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 safeguards programs.

Figure 2-2 illustrates the general structure of the CFFF organization.

2.1.2 POSITIONS AND ACTIVITIES WITHIN ORGANIZATIONAL OPERATING UNITS Each Westinghouse management position is covered by a written description, presenting in detad its scope, purpose, duties, responsibilities, difficulties, and requirements. The description identifies the incumbent's authority for decisions which may be made unilaterally, and those requiring higher management approval. It delineates relationships with other functions, and specifies responsibilities for managing personnel, and for the control and maintenance of managed facilities and equipment. Position descriptions are i

reviewed and approved by two higher levels ofline management. A Management Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

2.0 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

FIGURE 2-1 COMPANY ORGANIZATION WESTINGHOUSE ELECTRIC COMPANY (WEC)

COMMERCIAL NUCLEAR FUEL DIVISION (CNFD)

COLUMBIA FUEL FABRICATION FACILITY (CFFF)

Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

2.1 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

e.

FIGURE 2-2 CFFF ORGANIZATION 1

l l

l l

PLANT MANAGER i

ENGINEERING MANUFACTURING REGULATORY QUALITY COMPONENT COMPONENT COMPONENT COMPONENT l

i l ca==gs,y,a== p q on-1

.,o cr.. l

- i

.u-a ru.a - l 1,.mm,...

.. o - -a

==-l 1 ~~a-rua-l UNCTION

-4 ~w: arc" l 1

I 1

l

[==o=<aviggg*

""' 'c^'#,7

[- d *"'"fuic'"o!"

l l

'"v'a"r^,2"'

l

"*"*" A^,Z" l

l ""?!rE*T#cWo'n' l u

l l

Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

2.2 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

p e

i 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, 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 nified in the position description. Two higher levels of management, at a minimum, must approve each selection or change of a management incumbent. Continuing quality performance of managers is assured through a formal program of annual review.

~ Operations at the Columbia Fuel Fabrication Facility are in accordance with the general I

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 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 control, and to provide measurements, determinations and information which aid in the analysis of specific operations and situations; however, such service and staff assistance 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 makmg decisions at the operating level rests with the first level manager. A first level manager has the basic responsibility for operating controlled activities in a safe and prudent manner.

First level managers are responsible for pmviding operating instructions for the guidance 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 radiation safety rules and policies, use of protective clothing and personnel monitoring 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 Federal Regulations, are covered. Prunary emphasis is placed upon 10 CFR Parts 19 and

20. The cognizant first level manager assigns an experienced employee the responsibility ofindoctrinating 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 the I

l job description are fulfilled. Failure to achieve minimum performance requirements is i

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

2.3 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

t

~

l cause for a change in assignment, or for release from employment. Periodic reinforcement l

l 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 admmistrative 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 instmet their personnel regarding any remedial actions required.

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 authority to prohibit, through the cognizant first level manager, any operation which is believed tc involve undue unmediate hazard. Such termmated 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 ACCOUNTABIIJTY 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.

(a)

Plant Manager The Plant Manager will 'have overall accountability for all nuclear fuel manufacturing activities at the Columbia Fuel Fabrication Facility. This 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 managemerit experience in a nuclear facility. The Plant Manager will have broad general knowledge concerning the regulatory aspects of policies and procedures in effect at the Columbia Fuel Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

2.4 License No. SNM-1107

. Revision Submittal Date: 16AUG99 Revision No. 16.0

Fabrication Facility.

(b)

Component Managers Four Component Managers will have specific accountability for engineering, manufacturing, regulatory, and quality operations and activities involving licensed 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 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, quality, and safeguards, in all areas over which they have authority.

First Level Managers will supervise operating personnel. They will fulfill their responsibilities by assuring that all operations under their control are carried out in accordance with the radiation protection limits, nuclear criticality safety controls, processing procedures, schedules, and other instructions supplied by higher management.

All Component Managers will be knowledgeable in the operating procedures 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.

The minimum requirements for a Position of Component Manager, above the First Level, will be a baccalaureate degree, or equivalent, with a science or 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 Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

2.5

' License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

quality work accomplishments.

(c)

Regulatory Component Managers and Engineering Functions l

The Regulatory Component will be that organizational component of the licensed activity with the responsibility for environmental pollution control, radiation i.

protection, nuclear criticality safety, nen==tional safety and health, and emergency planning; and, for evaluating the effectiveness of these programs. The Regulatory l

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 have been evaluated and communicated to otMr Component management for incorporation into facilities, equipment, and procedures prior to their use for processing licensed material.

1 The Regulatory Component will, to the extent practicable, be administratively independent of manufacturing process supervision. The Regulatory Component will be responsible for the establishment, conduct, and continuing evaluation of licensed programs to ensure the protection of the employees at the licensed facility, of the public, and of the environment. In particular, for any processing change which could result in a credible consequence not previously evaluated, or in excess of one previously evaluated, the Regulatory Component will perform a 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 miministered 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 document the program's activities; and a program to maintain the above parameters As Low As Reasonably Achievable (ALARA).

l-Nuclear criticality safety services provided by the Regulatory Component will

. include as a minimum: the performance of process or equipment nuclear criticality

'~

safety analyses and evaluations before a new or modified fissile material operation is begun to include the determination of parametric controls and spacing requirements based upon validated analytical or computational techniques, l-including computation of effective neutron multiplication factors for fuel configurations; provision of audits, inspection and surveillance services to protect Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

2.6 j

Licanae No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0 j

~

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, chemW1, and fire hazards; the development and implementation of safety programs and. procedures ' designed to minimize accidents and injuries to 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 following:

License and permit administration; Routine surveillance of operations;

. Audits of licensed activities for compliance with applicable State and Federal regulations, licenses, and permits; and, docunaentation of these audits anxi 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, documentation of said conformance; Review of nuclear criticality safety, radiation protection, and occupational safety and health aspects of changes to equipment and operations associated j

with the processing, handling, or storage of nuclear material; Training in, and monitoring the training effectiveness of, environmental protection, radiation safety, nuclear criticality safety, occupational safety and health, and emergency planning; and, Monitoring, and reporting the effectiveness, of the program to assure radioactivity in effluents and radiation exposures are kept As low As Docket No. 70-1151' Initial Submittal Date:

30APR90 Page No.

2.7 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

r 1

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.

The minimum requimments for a position of Regulatory Function Engineer will be a hacenlaureate degree, or equivalent, with science or engineering emphasis; and, two years of nuclear industry experience in the assigned function. A Regulatory Function Engineer will have demonstrated proficiency in quality administration of

' the assigned position programs.

2.2 SAFETY COMMITTEES 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 pmgrams; 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 the Manager of the Regulatory Component, and at least three other Component Managers who are qualified to evaluate plant operations from a regulatory and safety standpoint. 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 Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

2.8 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

p-l 1

involvement, or when otherwise warranted by instant circumstances. The committee's findings, conclusions and recommendations will be formally documented to the Plant 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<ommittees.

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

l l

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

2.9 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No. 16.0

o..

i CH WrER 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 Improvement.

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 l

commitments will be followed to management-approved completion by the responsible individual j

specifically assigned accountability for each particular initiative.

This program will be a documented demonstration' of CFFF Managements' strong commitment to evaluate, on a i

continuing basis, opportunities to improve the Plant margin of safety - with the understanding l

that: addition, change, and/or deletion of program elements and/or initiatives; continuation of l

ongoing program elements. and/or initiatives; and/or, additions, deletions and/or changes of l

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 l

l To assure that design changes will not adversely impact on environmental protection, health, safety, quality, and/or safeguards programs at the Columbia Fuel Fabrication Facility (CFFF), a formal review process will be established to analyze new systems and components, or modifications to existing systems and components, in order to reliably predict performance under normal operating conditions and potential process upsets.

Structured hazard analyses, as conducted in accordance with Chapter 4.0 of this License Application, will specifically include analysis of verified drawings under configuration management.

3.1.1 ' CONFIGURATION MANAGEMENT PROGRAM AND PROCEDURE The CFFF Configuration Management Program will embrace an approved procedure for implementation of 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 requimnents of baseline safety assessments. The Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.0 License No. SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No.16.0

E

]

. procedure will specify documentation required to maintain a current record of existing system conditions.

3.1.2 CONFIGURATION MANAGEMENT IMPLEMENTATION

' 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 =mance 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 i

completing a configuration change control form, arxl enclosing specified project information for the review process.

(b)

Manufacturing, Engineering, And Quality Component Reviews Designated Manufactusig, 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.

i

'(c)

Regulatory Component Reviews For Approval 1

' 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 compliance with applicable regulatory requirements and conformance to regulatory commitments. Formal approvals will be documented as part of the review package.

(d)-

Ancillary Programs and Procedures Ancillary programs and procedures will be activated commensurate with identification of environmental protection, health, safety, and/or safeguards issues.

l Such programs will range from simple design reviews by cognizant multi-discipline Functions, through structured What-If/ Checklist or Hazards and L

Docket No.-

70-1151 Initial Submittal Date:

30APR90 Page No.

3.1 L

I-icense No. SNM-1107

' Revision Submittal Date: 16AUG99 Revision No.16.0 p

L.

-o Operability Analyses. Formal approvals will be documented as part of the review L

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.

l (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 sigidng the configuration change control form, attesting that all required documentation has l

been updated, all required training has been courpleted, and the project has been l

2mhW.

I 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 l

a condition of readiness such that it is likely to perfonn its desired function when called upon to do so.

The maintenance program will embrace three functional activities:

Progriutuued Maintenance, to include specified 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 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.

The following safety-related systems and components will receive programmed maintenance:

Air Compressors; Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.2 License No.- SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 i

L.

Emergency Electrical Generators; e

Fire Detection and Fire Control; e

. Natural Gas Valves; Nuclear Criticality Detection; e

Pressure Relief Valves; Steam Boilers.

Additional safety-mlated systems and components will be placed under programmed maintenance, as disclosed by the results of Integrated Safety Assessments described in 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 identified 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 discretion of the cognizant Regulatory Engineer Function.

Programmed maintenance of safety-related systems and components will include specified f

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, e

one day per working week; Plant-wide Hazard Warning System - Semiannual; i

Specified Safety-related Interlocks on Process Equipment - Annual; e

- Hydrogen and Natural Gas Line Leak Tests - Annual.

e 4

Additional safety-related systems and components will be placed under periodic functional testing, based on the results of integrated safety assessments described in 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 Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

3.3 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

FE L

contrcls identified through configuration management reviews of modifications of the l

system, will be scheduled, as eury, 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.

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 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 that such

~

equipment is designed, installed, operated, and maintained so that it will 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) program (29 CFR 1910.119), structured to include licensed radioactive materials. The Engineering Component will maintain a detailed matrix that graphically demonstrates how the PSM program elements will address the following QA program criteria:

(a)

QA Organization; j

(b)

- QA Program; Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No., 3.4 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

m

)

i l,

(c)-

Equipment / System Design Control; I'

(d)

Procurement Documentation Control; (e) '

Instructions, Procedures, and Drawings;

- (f)

Document Control; (g)

Contml of Purchased Materials, Equipment, and Services; (h)-

Identification and Control of Materials, Parts, and Components; j

(i)

Control of Special Processes; (j)

Internal Inspections;.

j (k)

Test Control; (1)

Control of Measuring and Test Equipment; (m)

Handling, Storage, and Shipping Controls;

.(n)

Inspection, Test, and Operating Status; i

(o)

Control of Nonconforming Materials, Parts, or Components; i

(p),

Corrective action; (q)

QA Records; and, (r)

Audits.

The PSM program will then be supplemented, as required, to assure detailed inclusion of all QA criteria.

3.3.2 GRADED APPROACH

- The ' graded approach" will be addressed by performing a systematic and integrated ass % ment of the hazards at the' facility; then, identifying the safety systems and components that are intended to prevent, or mitigate the consequences of, these hazards; then, to apply the~ programs of assurance which provide the appropriate level of quality.

(Completion of these assessments, as an ancillary supporting process, will be phased-in Docket No.-

70 1151 Initial Submittal Date:

30APR90 Page No.

3.5 License No. SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No.16.0

e 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 defined by controlling the following hazard consequences:

Greater than or equal to 5 rem dose equivalent to an individual offsite; and/or, Greater than or equal to 10 milligrams soluble Uranium intake by an o

individual offsite; and/or, Greater than or equal to 25 milligrams HF/m' exposure to an individual 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; Important Safety Systems i

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; e

Death or serious injury to an individual onsite.

e 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 l

' These systems have safety implications, but are neither crucial nor important to safety.

They do not require specified attention to quality assurance, and no extraordinary level of safety detail is applied. Safety margin improvement systems l

- will be maintained and operated as part of routine and prudent industry practice.

l Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.6 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

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.

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 with program requirements.

The program will include a level of checks and halance 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.

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-fitting upon implementation.

It will be a bounding assumption that existing systems were appropriately designed, installed, and operated in 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.

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 Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

3.7 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

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 will be integrated into a combined process to assure that environmental protection, health, integrated safety, quality, and safeguards programs are being conducted in accordance with Westinghouse policies, and in accordance with commitments to Regulatory Agencies.

Elements of this integrated process will be developed by knowledgeable Component staff, will be reviewed and approved by cognizant individuals in affected Components, and will be authorized for implementation by Component Management at a 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:

(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 i

safeguards traimng and instructions for Regulatory Functions.

They will be prepared, and approved for issuing, by Regulatory Functions assigned by a Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.8 License No. SNM 1107 Revision Submittal Date: 16AUG99 Revision No.16.0

cognizant Regulatory Component Manager; and, will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.

The Category-1 scope will group sets of procedures into such subcategories as:

Administration; Health Physics:

e Nuclear Criticality Safety Environmental Protection Safeguards e

Shipment and Transportation; Instruments; j

e Surveys; e

Dosimetry; J

e Bioassay; and, e

Laboratory Practices Changes to Category-1 Procedures will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.

(a.2) Category-2 Procedures 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 training and instructions for Engineering, Manufacturing, and Quality Functions; and, for use by these Functions in preparing Category-3 Procedures. They will present regulatory guidance methodology acceptable to the Regulatory Component. They will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and, will be reviewed, and approved

- for issuing, by the cognizant Regulatory Component 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.

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 reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.9 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

(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 - 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 cogmzant 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 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 quality assurance procedures.

I Regulatory-significant practices in operations and quality assurance procedures, and changes to such procedures, will be issued by cognizant Components in accordance with documented policies for procedure preparation, review, and approval.

Specifically, Regulatory Component approvals will be required for all regulatory aspects of procedures, and their changes, involving the storage, handling, processing, inspection, and/or transport of nuclear materials.

Component Management will be responsible for assuring and documenting that contents of these procedures are communicated to appropriate personnel through training programs, access to the Electronic Systems, and/or posting of instmetions.

(c)

. Procedure Review Frequencies Maximum frequencies of reviews-for-updating for regulatory-significant procedures will be:

i Docket No. 70-1151' Initial Submittal Date:

30APR90 Page No.

3.10 License No. SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No.16.0 1-

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 inadequate 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. 'Ihis 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 acknowledgment described in Paragraph 3.4.1(d) of this Chapter, and training responsibilities described in Chapter 2.0 of this License Application. Such traming 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 stmetured such that specified training and qualification requirements will be met prior to safety-significant positions being fully emed, or covered tasks being independently performed.

Training records will be maintained in accordance with Section 3.8 of this Chapter.

(a)

General, Topical, and Refresher Training All new employees 'will receive training in emergency response policies and guidelines, and general safety and regulatory practices. All new employees designated as radiation workers will receive additional 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, Docket No.-

70-1151 Initial Submittal Date:

30APR90 Page No.

3.11 License No. SNM-1107 '

devision Submittal Date: 16AUG99 Revision No.16.0

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 Iow As Reasonably Achievable (ALARA), and material safeguards. ' Facility visitors will be provided with traimng 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.

Radiation workers will receive regulatory refresher training on a biennial basis.

This training will consist of:

Providing each employee with a current revision of the Integrated Safety Traimng Manual; Piwniing each employee supplementary electronic instruction on general e

regulatory issues; and, Requiring each employee 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; e

General nuclear criticality safety practices; e

Area-specific nuclear criticality safety practices; e

Industrial safety and hygiene, and fire safety, practices; l

e Chemical Area work practices; e

Radiation risks; e

Emergency planning; and, Safeguards.

l Employees who are absent from the facility during scheduled regulatory refresher training will receive such training within one month of their return to work.

i

(

(b)

Traimng and Qualification of Nuclear Criticality Safety Function Engineers Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

3.12 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 1

j

L Nuclear Criticality Safety Function Engineers will develop skills and abilities l'

directed by the cognizant Regulatory Component Manager, who will evaluate furvhmental development methodologies for applicability and utilization on a case-i by-cases basis. Examples of development methods include:

A nuclear criticality safety short course; e

Westinghouse auditing certification; l

l-American Nuclear Society Standards development and review; Facility criticality safety handbook development and review; e

A structured hazards analysis course; e

e

-A structured human factors course; and, 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 include:

Repons of internal audits and inspections conducted; Feedback from worker training presented; e

Criticality safety analyses and evaluations performed.

Qualification of each Nuclear Criticality Safety Function Engineer will be formally l

documented by the cognizant Regulatory Component Manager and the senior l

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 include, as a minimum, a high school diploma or equivalent.

Health Physics Technicians will develop skills 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; e

Emergency preparedness training; and/or e

e' Applicable skills competency training.

Docket No.

70-1151 Initial Submittal Date:

30APR90 Page No.

3.13

. License No SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 i

3.5 HUMAN FACTORS Human factors concepts will be employed at the Columbia Fuel Fabrication Facility (CFFF), in recognition of how the total job envirorunent - areas, equipment, training, and procedures - shapes the expectations, thoughts, arxi decisions of employees who work with licensed 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 emixxlied in procedures, as described in Subsection 3.4.1 of this Chapter, procedures will receive particular human factors attention.

3.5.1 DEVELOPMENT OF HUMAN FACTORS AWARENESS i.

1 To enable integration of human factors concepts into facility operations, an initial, formal course - prepared and presented by recognized human factors experts - will be provided for the Plant Manager; all Engineering, Manufacturing, Regulatory, and Quality i

Component Managers; and, designated Functions from these Compotents. The course will address the following elements, including exercises to enhance learned skills:

(a)

Process Safety Management; (b)

Human Factors Concepts; (c)

Performance Shaping Factors For Hardware; j

(d)

Performance Shaping Factors For Procedures; 1

l (e)

Analysis Preparation; i

(f)

Error-Likely Situations; (g)

Procedure Analysis Techniques; 1

(h)

Worker Self-Checking Techniques; and, (i)

Supervisor Coaching Principles.

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 l

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.14 License No.' SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

m I

^

l 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 Structured Analysis Of Procedures.

(b)

Minimiring' Opportunities For Human Errors Of Omission and Commission Related To Procedures.

Results. of the stmetured 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 descrioed 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)

INFORMAL INSPECTIONS Regulatory-Component personnel on duty, including Regulatory Component management, will conduct continuing informal inspections of regulatory program i

= performance in the course of their routine duties. Observed process upsets and I

procedural. inadequacies will be promptly reported to the cognizant First Level Component Manager for remedial action. Repeated upsets and inadequacies will l

be reported to the cognizant Regulatory Component Manager, who in turn will report them to increasingly higher levels of Component Management until effective remedial action has been taken. Such repeated upsets and ' adequacies will be m

Mmented in monthly formal audits to assure applicable tracking and resolutions.

(b)

FORMAL AUDITS

. Cognizant Regulatory Function Engineers will conduct monthly formal audits of regulatory pmgram performance in accordance with a written procedure. The auditors will have the technical capability, and will be formally directed by Regulatory Component management, to fmd process upsets and procedural l

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 Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.15 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

Component management, for the areas being audited; and, detailed area walkdowns. Disclosed upsets and inadequacies will be formally documented in a report to cognizant 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 Staffs 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 1

they self-reveal and/or Regulatory Agencies fmd them.

On a semi-annual basis the following assessment parameters will be summarized and trended by the Regulatory Component:

A summary ofitems documented in the performance-based reporting process; A summary of upsets and inadequacies documented in performance-based internal e

audit reports; Facility Collective Dose Equivalent; e

Facility average Total Effective Dose Equivalent; e

Top 10 facility workers' Total Effective Dose Equivalents; Overexposures; e

Regulatory Agency notifications; i

e Ratio of Recordable Incident Rate to SIC code average; j

e Lost time accidents per production hour; e

Results' of Special Nuclear Material Physical Inventory (annual);

Emergency response team activations; e

Radioactive emissions in gaseous effluents; e

Radioactive emissions in liquid effluents; e

Radioactive material transportation incidents; and, e

Regulatory Agency violations.

e Docket No.~

70-1151 Initial Submittal Date:

30APR90 Page No.

3.16 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

[

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.

3.7 INCIDENTINVESTIGATIONS i

At the Columbia Fuel Fabrication Facility (CFFF), the organizational structure 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 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 failure (s) that led to investigated events.

3.7.1 INTERNAL REPORTING OF INCIDENTS A formal system will be maintained to enable employees to report process upsets and procedure inadequacies to their First level Managers for follow-up action; and, employees will be instructed 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, e

Acknowledgement of closure (and date acknowledged) by the cognizant e

Engineering Function Engineer, the cognizant Regulatory Function Engineer, the originator's First level Manager, and the originator.

Potential safety-significant reports will be forwarded to the Regulatory Component for 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-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 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. Stmeturcd incident evaluations will be Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.17 License No. SNM-Il07 Revision Submittal Date: 16AUG99 Revision No.16.0

r 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 e

conducting such an analysis, including recommendations; Status of corrective action (s) implementation; Regulatory assessment; e

Notification documentation; e

Traming documentation; e

Plant-wide applicability assessment; and, I

e Miscellaneous information pertaining to the incident and/or the evaluation.

e 3.7.3 NOTIFICATION OF REGULATORY AGENCIES Cognizant Regulatory Agencies will be promptly notified of major safety incidents in accordance with all requiments from 10 CFR Parts 20 and 70. In particular, as points of l

additional clarification, the NRC Operations Center will be notified of the following types l

of incidents, within the time limits prescribed:

(a) 1-Hour Notifications (a.1) Any incident for which an Alert or Site Area Emergency has been declared, as prescribed by the Site Emergency Plan described in Chapter 9.0 of this License Application.

j

)

(a.2) Any incident involving Quality level A systems, for which accident controls j

cannot be initiated, whether or not regulatory limits are exceeded.

1 (b) 4-Hour Notifications (b.1) Any incident involving Quality level B systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded.

l (b.2) Any nuclear criticality safety incident for which less than double contingency protection remains (multi-parameter control or single-parameter control) and:

I Greater than a safe mass is involved and double contingency protection is e

l not restored within four (4) hours, Greater than a safe mass is involved and controls are restored within four e

(4) hours, but:

i. Only single contingency protection is restored and more than one of the original controls were modified or replaced.

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.18 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

ii. Double contingency protection is restored but multiple original controls l

under both contingencies were modified or replaced.

(b.3) Any determination that a criticality safety analysis or evaluation was deficient and that double contingency protection, in fact does not exist.

(b.4) Any unanticipated /unanalyzed nuclear criticality safety incident for which the severity and remedy are not readily determined.

(c) 24-Hour Notifications (c.1) Any incident for which the work area is unavailable for normal use for an entire day, following a loss of radioactivity contamination control.

)

i (c.2) Any ircident 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 removable radioactivity contamination 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.

(c.5) Any nuclear criticality safety incident for which less than double contingency protection remains (multi-parameter control or single-parameter control) and:

Less than a safe mass is involved.

Greater than a safe mass is involved, but a sufficient number of the controls i

e that were lost are restored within four (4) hours - such that double contingency protection is restored.

)

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.19 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 l

l

i

.(

3.8

'RECORDKEEPING AND REPORTING The Columbia Fuel Fabrication Facility will identify, mamtain, preserve, control, and destroy records - as defined in the records management section of the controller's manual

- in ' accordance with the gn*1im, procedures, and practices set fonh by the Westingham 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 applicable 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)-

Environmental Surveys;

- (b)

Radiation And Contamination Surveys;

.(c)

' Personnel Exposures; (d)

Instrument Calibration Results;

'(e) : Nuclear Criticality Safety Evaluations, Analyses and Methodology Validations; (f).

Audit And Inspection Reports; (g)-

. ALARA Reports;-.

(h)

Regulatory Compliance Committee Meeting Minutes; l

Docket No. 70-1151-Initial Submittal Date:

30APR90 Page No.

3.20 l

License No,' SNM-1107-Revision Submittal Date: 16AUG99 Revision No.' 16.0

)

I (i)

Employee Training And Re-training Documentation; (j)-

Records Of Plant Alterations Or Additions; (k)

Documentation Of Abnormal Or Atypical Occurrences And Events Associated With Radioactivity Releases; (1)

Decontamination And Decommissioning Files; and,

(

(m)

Other Such Records Required By the Regulations.

These procedures will include Records Flow Schedules, which list:

Record category, e

Name of record; Form numbers; e

Retention period in active files; j

Retention period in the central records bureau; and,

~

e Retention period in the records center.

e 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.

Records of nuclear criticality safety analyses will be retained for the lifetime of the facility.

3.8.2 RECORDS RETRIEVAL I

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.

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 destruction will not foreclose capability for reconstructing a complete and correct set of required records.

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No, 3.21 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

r o

In cases where protective measures fail, and a particular record is lost or inadvertently l

l 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 reconstructed using inference to data relative to other documents for shnilar information and time periods.

3.8.4 REPORTS A detailed listing of reports required by NRC regulations will be maintained and followed.

This listing will document:

Reference to applicable regulations; Descriptions of the reports required; and, Frequencies at which the reports must be submitted.

e Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

3.22 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 I

F CHAPTER 10.0 ENVIRONMENTAL PROTECTION 10.1 EFFLUENT AIR TREATMENT For operations that might result in exhausting radioactive materials to unrestricted areas, the adequacy of air emuent control will be determined by representative stack sampling l

to demonstrate compliance with the regulations.

Sampling will be performed l~

continuously during production operations. Samples will be collected and analyzed daily l

during production operations. If radioactivity in plant gaseous emuents exceeds 1,500 l

microcuries per calendar quarter, a report will be prepared and submitted to the NRC Staff within 30-days of the end of the quarter in which the incident occurred. This report will klentify the cause for exWag the limit, and corrective actions to reduce the release rates. The report will be submitted to NRC Headquarters Staff, with a copy to NRC Region II. Suhs~=wmtly, if any parameters important to a dose assessment in tne original report are found to have changed, a follow-up report will be submitted, within 30-days, which describes the changes in parameters and includes an estimate of the resultant change in dose commitment. In the event that a calculated Total Effective Dose Equivalent to any member of the public, in a calendar year, threatens to exceed 100 MREM per year, immediate steps will be taken to reduce emissions to levels that will assure compliance.

10.2 LIQUID WASTE TREATMENT FACILITIES A liquid waste treatment facility, with sufficient capacity and capability to enable holdup, treatment, sampling, analysis, and discharge of liquid wastes in accordance with the regulations, will be provided and maintained in proper operating condition.

Control of radioactivity in the ADU process liquid emuent waste stream will be achieved i

by the operation of two treatment systems: (1) a continuous on-line gamma spectroscopy

. monitor and quarantine tank filtration system, within the chemical controlled area, and (2) an advanced wastewater treatment facility to remove the last remnant of uranium, outside the facility.

The first system will be installed following quarantine tanks, diversion tanks, and filtration operations. 'Ihis system assures that the ADU process liquid waste emuent I

being discharged from the chemical controlled area to the external waste treatment facility meets discharge criteria established by plant operating procedures, nominally less

. than 30 ppm uranium (equivalent to 7.2 E-05 uCi/ml at a specific activity of 2.4 l

Docket No. ' 70-1151 Initial Submittal Date:

30APR90 Page No.

10.0 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

uCi/gU). When the liquid has been successfully anna ~i for discharge, it will be pumped from the inplant final pump out tank to the second system, the advanced waste water treatment facility for uranium removal external to the main plant.

This second advanced wastewater treatment system will assure that the last remnant of uranium in the discharges is removed from the process liquid stream to a nominal limit of less than 0.5 ppm uranium (equivalent to 1.2 E-06 uCi/ml at a specific activity of 2.4 uCi/gU). Established p*. ant opining procedures will assure that NRC 10CFR20 liquid discharge limitations are met, and will implement ALARA control.

Other miscellaneous liquid waste will be filtered and sampled on a batch basis to assure uranium is effectively removed to levels which will enable conformance to ALARA

- goals. Quiescent settling in lagoons (East, West, North, and South) will further enable uranium removal to levels which will assure contianing compliance with 10 CFR 20.

1301 and 1302 limits.

A continuous, proportional sample of liquid effluent released to the Congaree River will be collected. A 304ay composite of this sample will be analyzed for gross alpha activity, gmss beta activity, and isotopic uranium content.

Any violation of the facility NPDES Permit will be reported to NRC Region II within 15-days of confirmation of the violation. If the NPDES permit conditions are revised, or if the permit is revoked, the NRC Headquarters Licensing Staff will be promptly notified.

10.3 SOLID WASTE DISPOSAL FACILITIES Solid waste disposal facilities, with sufficient capability to enable preparation, packaging, and transfers to licensed disposal sites in accordance with the regulations, will be provided and maintained in proper operating condition.

10.4-PROGRAM DOCUMENTATION

'Ibe licensed activity prepared an Environmental Evaluation Report dated March 1975, that has been anha-amtly updated in revisions dated April 1983 and April 1990. Future i

Environmental Impact Appraisal updates will be prepared and submitted to the NRC Licensing Staff on a schedule contingent upon the ogiatirg term of the license. For a 10-year license, the review will be dmunmted in the ALARA Report (described in Chapter 5.0 of this License Application) and uping will be concurrent with each renewal application.' The substance and methodology of each such update will be as 4

agreed upon by cognizant NRC Licensing Staff and representatives of the licensed activity.

Docket No. 70-1151' Initial Submittal Date:

30APR90 Page No.

10.1 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

I 10.4.1 MINIMUM PROGRAM IMPLEMENTATION The Columbia Fuel Fabrication Facility environmental monitoring program will include the elements illustrated in Figure 10-1. For wells found not to contain water at time of l

sampling, an evaluation will be petformed by the Regulatory Component to determine if alternate well data may be used or a new well must be dug. Minimum program analytical sensitivities will be as illustrated in Figure 10-2. lax:ations of air, vegetation, and soil monitoring stations, locations of surface water monitoring stations, and locations of monitoring wells, will be as illustrated in Figures 10-3,10-4, and 10-5, respectively.

Action levels will be established by procedure for environmental samples. These program elements, analytical sensitivities, and/or locations may be changed without prior NRC Licensing Staff approval, provided:

(1) a documented evaluation by the Environmental Protection Function demonstrates that the changes will not decrease the overall effectiveness of the environmental amnitoring program; and, (2) the changes and d==*i evaluation are submitted to the NRC Licensing Staff as part of the subsequent Environmental Impact Appraisal update.

10.4.2 REPORTING PROGRAM RESULTS Radioactivity in releases of radioactive materials in gaseous and liquid effluents from the facility will be reported to the NRC Staff, in accordance with the regulations and applicable Regulatory Guide documents, on a semiannual basis.

10.5 EVALUATIONS The Regulatory Component will perform a biennial evaluation of vendors used to analyze environmental samples. Such evaluations will also be performed if substantive program I

anomalies are disclosed.

The evaluations will consider the need for " spike" and

" replicate sample" submittals.

10.6 OFF-SITE DOSE Compliance with NRC 10 CFR 20, Subpart D and EPA 40 CFR 190 regulations for off-site dose requirements to the maximally exposed individual will be demonstrated by assuring that the off-site annual dose does not exceed 25 MREM. The calculational methodology 'will include models which have been evaluated by the Regulatory Component and are deemed acceptable by the appropriate regulatory agencies.

1 i

l Docket No. 70 1151 Initial Submittal Date:

30APR90 Page No.

10.2 l

License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

FIGURE 10-1 CFFF ENVIRONMENTAL MONITORING PARAMETERS TYPE OF SAMPLE LOCATIONS ANALYSES SAMPLING FREQUENCY Air Particulates Four Alpha Continuous (Collection Weekly)

Surface Water "Ihree Alpha; Beta Quarterly Well Water' Ten Alpha; Beta; Ammonia Quarterly River Water Three Alpha Quarterly Sediment One Alpha; Beta; Uranium Annually Soil Four Alpha; Beta; Uranium Annually Vegetation' Four Alpha; Beta; Fluoride Annually Fish One Alpha: Beta; Uranium Annually FIGURE 10 2

'If gross alpha concentration exceeds 15 pCi/1, isotopic analyses for uranium will be conducted.

If gross beta exceeds 50 pCi/1, isotopic analyses for beta will be performed. If a monitoring well exceeds a mean concentration of 30 pCi/l of total uranium, the result will be provided to cognizant NRC staff.

'If a vegetation gross alpha activity result exceeds 15 pCilgram an additional sample will be collected.

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

10.3 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0 j

ENVIRONMENTAL MONITORING PROGRAM SENSITIVITIES TYPE OF ANALYSES TYPICAL NOMINAL MINIMUM SAMPLE QUANTITY DETECTION LEVEL Air Particulates Alpha 571 Cubic Meters 2.0E-15 Microcunes Per Milliliter Surface Water Alpha 1 Liter 2.2E-9 Microcuries Per Milliliter Beta 1 Liter 2.5E-8 Microcuries Per Milliliter Well Water Alpha 1 Liter 2.2E-9 Microcuries Per Milliliter Beta 1 Liter 2.5E-8 Microcuries Per Milliliter River Water Alpha 1 Liter 2.2E-9 Microcuries Per Milliliter Beta 1 Liter 2.5E 8 Microcuries Per Milliliter Sediment Alpha 100 Grams 1.0 Pacocurie Per Gram Beta 100 Grams 3.0 Picocuries Per Gram Uranium 100 Grams 0.5 Picocunes Per Gram Soil Alpha 100 Grams 1.0 Picocurie Per Gram Beta 100 Grams 3.0 Pacocunes Per Gram Uranium 100 Grams 0.5 Picoeuries Per Gram Vegetation Alpha 100 Grams 3.0 Pacocuries Per Gram l

Beta 100 Grams 0.5 Picocunes Per Gram Fish Alpha 30 Grams 1.0 Picocune Per Gram l

Beta 30 Grams 3.0 Picoeuries Per Gram Uranium 1 Kilogram 0.5 Picrocuries Per Gram FIGURE 10-3 Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

10.4 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

FIGURE 10-3 AIR, VEGETATION, i.ND SOIL MONITORING LOCATIONS

/

-z-kN

.gj

• .G.

.,e.

s l

l I

y f

7s s

5 l

!s i

\\

d

=

fI ij

• i. fi N.,Q *

\\') l(;j

/

j

\\it 'T

,\\

y AN s

(

/n i

(

11 N

/,

i 4,

k

,Mh.Y [

/ !!

j. ~ 1 i

\\ it' k

\\,

I{;M/

.d

.I' i.l.'

l 1 '/ J

^

, y

/

o f;hp[.$. l-97

/

I.'s j, 'N ty., '(p,

.f f

kl

.5:

/

• i

'. h.

/

ai

\\

=

s b.

m i5 g i

\\.

\\

f'l g i il til

\\

-1 O4O l

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

10.5 License No. SNM-Il07 Revision Submittal Date: 16AUG99 Revision No.16.0 r

.i--.

FIGURE 10-4 SURFACE WATER MONITORING LOCATIONS W\\

i

~.s # j m c,%

~, - - -

/

..h.

{y_r

/. M

.T

.n.

,c.

(

(

(.o

\\.-)4

~.... ;y j

\\/

~.....

~;-

~ ~.........(;: -

., N...

_. / - ca me

..\\;,,,'f/ M,

.e l

i

/

. 'Y ".

,,.. d e ps'l.

' ' ' ~ ~. _,

$~ %e

~ ~ ~ ~..

n - (.

.. - ~

.p.

(

i

{

180T TO SCALE

• • • h N

IAT scossoas sT. enig

[

COLuassea,3.C.

/

Y

'w$..r.%' %.

[

\\

2.aev

?

)

h.

• O u.c nowrom stato.s T.

• g,

??....

Je i

rg;as# r-y_/

scats orantes Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

10.6

~

License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

FIGURE 10-5 GROUND WATER MONITORING LOCATIONS

,I W 33

.7,*

}g LAcOONS

'g t," #.

wi s

I* "

SANIT.

,,g;. w 26

'*\\

t. coo.

e g..

t.h.e c

W-7 g

S. S g,

A.

s 4,.

' ' jig 2p,

, e,',,

W 32

• • .* S y k

SLMSET I...,

'e..,

"E w i3 w-se,:,/,.',,*,4

.w,2b-

}

E *,:f

\\

\\/

/J # *

• W 3A

?*

h. A,

....\\

\\.,. =

s:-

' i:f.,

w in f,g ;; 'f J.*

W 23

.e J d8 g

'G !*-l ji e;,.l'o

~

see e

SCJL1 et FEET LEGEND

'+/

e MONITORING WELL LOCATION

/-

WASTEWATER TREATMENT LAGOON Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

10.7 License No. SNM-1107 Revision Submittal Date: 16AUG99 Revision No.16.0

3.-

CHAPTER 11.0 DECOMMISSIONING To assure adequate financial resources will be available to decommission the Columbia Fuel Fabrication Facility (CFFF) at the end of its useful life, a conceptual deconunissioning plan

(" COST ESTIMATE TO TERMINATE LICENSE SNM-1107"), and a decommissioning funding

. plan a:yl financial assurance mechanism will be prepared and maintained current.

11.1.

CONCEPTUAL DECOMMISSIONING PLAN In suppod'of the~ COST ESTIMATE TO TERMINATE LICENSE SNM-1107, a document file will be maintained.

This file will consist of the following record categories:

(a)

Correspondence Chronological File; (b)- Historic Conceptual Plan (s) and Cost Estimate (s);

(c)

Historic Facility Radiological Information; (d)

NRC Guidance Documents;

..(e)

EPA Guidance Documents; (f)- Decommissioning Plan Shell; (g)

Current Conceptual Plan And Cost Estimate; and, (h)

Financial Assurance.

The file will include a record log-out/ return process that provides for information on:

"date", "out to", and " file number or name out"; and, each record category will be (clearly marked: " warning, these decommissioning records must not be removed or destroyed without the written approval of(the Regulatory Component)".

" Copies of the most recent COST ESTIMATE TO TERMINATE LICENSE SNM-1107 will be maintained by 'the Regulatory Component and/or the Engineering Component.

The Engineering Component will maintain the electronic copy that contains the Westinghouse position in the following file structure:

1 Docket No. 70-1151 Initial Submittal Date:

30APR90 ' Page No.

11.0 License No. SNM-1107 -

Revision Submittal Date: 16AUG99 Revision No.16.0 o-

.,:.i*'*

e L

- (a) ' ' Executive Summary;

-(b). Project Summary; i

L (c)1 Project Description; (d)L Estimate Configuration; (e)

Assumptions;.'

(f)

Westinghouse Staff; (g)

Demolition Labor Rate; (h). Subcontract - Consumables; (i)

Wash-Down Estimate;

]

(j)

Labor Factors; (k) ' Material Density And Pack Factors; (1). ' Inflation Factors; (m) Major Cost Drivers; (n)

Overhead Piping Density; J

(o) - Structure Data Sheets;

.(p)

Equipment Data Sheets; and, (q)

Major Drivers.

1 The COST ESTIMATE TO TERMINATE LICENSE SNM-1107 will be reviewed for need to update on a triennial basis.

11.2 :

DECOMMISSIONING FUNDING PLAN' AND FINANCIAL ASSURANCE MECHANISM i

To substantiate the cost total for decommissioning, the Westinghouse position on the following cost estimating tables will be maintained.

I

! Docket No.'

70-1151

-Initial Submittal Date:

30APR90 Page No.

11.1 l

License No.= SNM-1107-

' Revision Submittal Date: 16AUG99 Revision No.16.0

(a)

Planning And Preparation;

'(b)

Decontamination And/Or Dismantling Of Radioactive Facility Components; (c) - Packaging, Shipping, And Disposal Of Radioactive Wastes; (d)

Restoration Of Contaminated Areas On Facility Ground; (e)

Final Radiation Survey; and, (f)

Site Stabilization, IAng-Term Surveillance.

The Westinghouse Electric Company has established a Decommissioning Funding Plan including the necessary Financial Assurance Mechanism in accordance with the provisions of 10CFR70.25. The latest revision to the Decommissioning Cost Estimate for License Number SNM-1107 was submitted by Westinghouse letter dated August 29, 1997.- Revised financial assurance instruments to reflect the revised cost estimate were transmitted by Westinghouse letter dated July 10, 1998.

These revisions were acknowledged and accepted by the NRC by letter dated July 23,1998.

By letters dated September 28, 1998, November 16, 1998, January 18,1999 and February 22,1999. Westinghouse requested the transfer of License Number SNM-1107 from CBS Corporation to Westinghouse Electric Company LLC in conjunction with the sale of the assets to the nuclear and government operations business of CBS Corporation to'a consortium consisting of Morrison Knudsen Corporation and BNFL USA Group, Inc.-

In conjunction with the transfer of the license, revised f' ancial assurance m

documents were submitted to the USNRC by letter dated March 30,1999 and further amended by letter dated May 18, 1999. These latter two submittals provide the most recent financial assurance mechanisms for the Decommissioning Funding Plan for License Number SNM-1107 of the Westinghouse Electric Company. By letter dated e August 3,1999 the USNRC accepted the revised financial assurance documents.

Future updates of the decommissioning cost estimate and related revisions to the financial assurance mechanisms, will be provided in accordance with the prevailing license conditions and /or regulator directives.

Docket No. 70-1151 Initial Submittal Date:

30APR90 Page No.

11.2 License No.- SNM-1107

Revision Submittal Date: 16AUG99 Revision No.16.0 1

L.