ML19350C331

Demonstration D-II-2:Safety Analysis, Section 15

ML19350C331
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Site:	07002909
Issue date:	12/22/1980
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML19350C329	List: ML19350C328 ML19350C331 ML19350C333 ML19350C336 ML19350C339
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18190, NUDOCS 8103310937
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_

O SECTION 15 Q

DEMONSTRATION 0-11-2: SAFETY ANALYSIS Paragraphs 70.22(a)(1), (2), (3), (4), (6), and (7) of 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material" state that an application for a license must contain information v'

relating to:

Applicant identification m

a General activity plans a

Term of the license Special nuclear material parameters a

Technical qualifications of key applicant staff a

A description of equipment, facilities, and procedures (administrative controls) to be a

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used to protect health and minimize danger to life or prcperty.

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The following safety analysis is presented in support of this license application.

15-1 GENER AL LICENSED ACTIVITY INFORMATION Westinghouse corporate informatien was originally submitted to the U.S. Nuclear Regulatory Commission in a letter addressed to Mr. R. W. Lowenstein, Assistant Director of Regulations, dated April 3,1964, and thereafter has been (and will continue to be) updated approximately l

annually. The last previous letter, dated September 4,1980, was transmitted jointly to Mr.

C. V. Smith, Jr. (Director, Office of Nuclear Material Safety & Safeguards), and to O

Mr. H. Denton (Director, Office of Nuclear Reactor Regulation).

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15-1.1 Corps. ate Information i

The Westinghouse Electric Corporation is incorporated in the Commonwealth of Pennsylvania, with principal offices located in the Westing' ouse Building, Gateway Center, Pittsburgh, Penn-n l

sylvania 15222.

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18130 D-3.1 l

8103310937 e

Westinghouse is a publicly held corporation whose stock is traded on principal securities ex-changes. It is not owned, nor is there (to the best of our knowledge) an appreciable ownership of Westinghouse stock, by an alien, foreign corporation or foreign government. No individual is known, from the records of the Corporation, to own one percent or more of its capital stock.

Enclosed with the referenced September 4,1980 letter was the annual report of the Corpora-tion, which gives the current financial condition and lists the elected officers.

15-1.2 Corporate Technical Qualification The Westinghouse Electric Corpc<ation has bud experience in the field of nuclear science and technology. The Corporation's participat.. in the nuclear energy field dates from the discovery of methods for the production of rc..allic uranium at Bloomfield, New Jersey, in the 1920's and construction of the first industrial Van de Graaf generator in Pittsburgh in 1937. Westing-house furnished a portion of the refined metallic uranium used in the first pile at Stagg Field, Chicago, early in the 1940's, at the beginning of the Manhattan District of the Corps of Engi-n:ers.

Westinghouse demonstrated its ability to execute complex programs leading to the practical application of nuclear energy with the successful completion of the reactor plant for the first nuclear-powered submarine, the U.S.S. Nautilus. In conjunction with this project, the Bettis Atomic Power Laboratory was organized in 1948 to furnish a research and development effort.

This laboratory, which provides facilities for developing nuclear power plants for naval and advanced civilian applications, is currently being operated by Westinghouse for the Department of Energy (DOE). Westinghouse also executed the contract for the desigr and construction of the nation's first large nuclear reactor plant for an electric power generating station, the Ship-pingport Atomic Power Station. Other projects include a minimum of 38 completed power reactors including the nuclear power plant for the Yankee Atomic Electric Company, a 185 MWe, first round demonstration, pressurized water reactor which has generated over twenty billion kilowatt-hours of electricity; the Saxton Reactor, a 23.5 MWe experimental closed cycle water reactor which provided valuable data and experience in the use of plutonium recycle fuel in a pressurized water reactor; the second-generation Connecticut Yankee Atomic Power Co.

plant, a 575 MWe pressurized water reactor which was the first of the large, economically competitive reactora; and the Zion 1 nuclear plant, which is the first operational single-reactor generating unit rated in excess of 1,000 MWe. These operating reactors have a total generating capacity of approximately 25,425 MWe. Currently, the Corporation is designing or building 70 additional large reactor facilities, ranging in size from 350 MWe to 1250 MWe, with a total generating capacity in excess of 71,679 MWe. Also, the fabrication of replacement regions of fu:1 for operating reactors is a significant activity.

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F During March,1979, the Westinghouse-designed Trojan plant generated a record 801,451,000 kwh

~ I net, with a capacity factor of 99.7 percent. The Trojan plant, a 1030 MWe, 4-loop Westing-I

. house PWR is focated on the Columbia River in Northern Oregon and is operated by the Port-land Genmal Electric Company.

4 The Corporation holds the contract to provide tha project management, design, and test serv-ices for' the Fast Flux Test Facility, which will be used in the testing and evaluation of fuels and materials for DOE's Liquid Metal Fast Breeder Reactor (LMFBR) program.

Westinghouse has been a leader in the development of nuclear propulsion and auxiliary power equipment for space applications. The former Westinghouse Astronuclear Laboratory developed

.and fabricated nuclear reactors for the NERVA program and participated in the development of the SNAP-23A pack:ge and compact thermoelectric converters for the U.S. Government.

Various divisions of the Corporation have demonstrated other major accomplishments in the nuclear energy field. Westinghouse sveloped canned motor and controlled leakage pumps, currently being manufactured for a variety of nuclear facilities, and it also manufactures many other non-nuclear components for reactor plants such as large heat exchangers, control rod.

drive mechanisms,: valves, instrumentation and control equipment.

Westinghouse maintains a number of design and development groups in the Pittsburgh area

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(over 3,600 engineers and scientists) that contribute to these accomplishments in the nuclear.

F field.' Thera is an' accident ' prevention administrator and a medical services administrator 10-cated at the Gateway. Center Headquarters in Pittsburgh. At another Westinghouse 1ocation

.near Pittsburgh,' there is a headquarters-industrial hygiene administrator whose engineering and c

U laboratory facilities are available to all locations. The headquarters staff for the Nuclear Energy 1

Systems (NES) group includes a Director of Safety and Industrial Hygiene, who conducts-lt

special projects, drafts general policies, and provides coordination among.the Industrial Hygiene supervisors'at the various NES sites; a. manager of License Administration for coordination of

. licensing activities; and a Manager of Nuclear Materials Management and Safeguards to provide guidance and advice on safeguarding special nuclear. materials. Each site performing nuclear 1 activities has 'at least one technically qualified,- full-time supervisor, with: additional engineers -

i and technicians as.neededl in support of radiation ' protection,' industrial h'ygi.ene, and safety;

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' services. FulFtime scientists and engineers.witt, extensive experience in nuclear design lend.

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support to the various facilities by providirig nuclear criticality safety analyses where special

' nuclear materials are used. Computer service is available'for determining safety parameters in :

q nuclear criticality safety: analyses..

i Facilities in'the.Pittsburgh area inclu& a wide variety of operations,.~ ranging from research and fdevelopment to' pilot-plant scale manufacturing,1which require handling and processing many.

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(types of radioactive. materials ranging in quantity from a few microcunes upito megacunes,

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Approximatdy 8700 employees are engaged in nuclear activities at these facilities, which occupy tbout 3,060,000 square feet of floor space. Additional facilities at other geographical locations involving approximately 6400 people and 1,800,000 square feet of floor space, greatly expand the Corporation':, capabilities in the nuclear energy field.

15-1.3 Current List of Westinghouse - U.S. Nuclear Regulatory Commission Licenses USERS AND SITE LICENSE NUMBERS Nuclear Energy Systems a

Nuclear Center SNM 1506 Cheswick, Pa.

SNM-1120, 37-05809-01, 37 04809-02 Columbia, S.C.

SNM 1107 Irigaray, Wyoming SUA-1204 Forest Hills, Pa.

37-00497-09, SNM-38 Uoi it)

Large, Pa.

SNM-951 Waltz Mill, Pa.

SNM-770, TR-2 Zion,111.

R-119 Copperton, Utah SUA-1315 m

Pesearch Laboratories Churchill, Pa.

SNM 47, SNM-1460 m

Semiconductor Division Youngwood, Pa.

37-07934 01 15-2 DETAILED OPERATIONS EVALUATION The Alabama Nuclear Fuel Fabrication Plant (ANFFP) Special Nuclear Materials (SNM) Build-ing is to house licensed material manufacturing operations. The SNM Building is to be basically constructed to criteria embodied in the Southern Standard Building Code, with appropriate special additions and exceptions related to the nature of the business to be conducted. The special additions are specified in Section 5 of this license application, and are further detailed in the following safety analysis. A major special addition is inclusion of a nuclear criticality safety moderation control water barrier" which totally encloses operations of conversion through blending, and consequently provides a supplementary building containment for tne bulk of facility materials (uranium oxide powders) in an air dispersible form. Special exceptions are O

D-3.4

also specified in Section 5 of this license application; these relate to location of fire safety

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sprinklers, as necessary to enhance nuclear criticality safety moderation control, and facility exits as necessary to enhance physical security. Facility designs are to be reviewed with Fac-tory Mutual and American Nuclear insurers consultants at appropriate times prior to, and dur-ing, facility construction.

The Manufacturing Department of the Westinghouse Nuclear Fuel Division (NFD) is primarily engaged in the manufacture of fuJ assembir~; for commercial nudear reactors. The manufactur-ing operations to be authorized by this license consist of receiving low-enriched (less than or equal to 5.0 w/o U-235) uranium hexafluoride (UF ); processing the hexafluoride to produce 6

.O uranium dioxide (UO ) powder; and processing the dioxide powder through pellet pressing and (j

2 sintering, fuel rod loading and sealing, and fuel assembly fabrication. These operations (depicted in block diagram form in figures 3-2 through 3-9 of the ANFFP Environmental Report) are accompanied by appropriate nuclear criticality safety, radiation protection, environmental, and quality assurance controls.

The uranium processing operations to be performed under this license are conveniently divided into five distinct categories. The first category is the IDR Conversion operations which employ chemical means to convert UF6 o UO2 powder; a second category is the Fabrication operations t

which use essentially mechanical processe: to produce fuel assemblies containing encapsulated UO2 pellets.

The third category includes Ancillary Services in support of manufacturing activities, especially laboratory operations which utilize a variety of spectrographic and wet chemical analyses of small samples of product materials to assure that product specifications are met. (The ANFFP analytical laboratory is part of Quality Control within the Product Assurance Department of NFD. This laboratory is to perform chemical analyses on products produced in manufacturing operations. It may also supply similar services.for other Westinghouse facilities or outside cus-tomers. The laboratory will possess and' use instrument calibration standards which include gram quantity samples of uranium containing U-235 at a variety of enrichments-up to and includ-(

ing fully. enriched - and gram quantity samples of U-233.)

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A fourth category is Scrap Recovery, the t'eatment of scrap generated by manufacturing activi-

. ties, to permit it to be either recycled back into production operations or to be more. closely controlled prior to discard as waste. Such treatment includes chemical dissolution and precipita-

' tion, and dry processes.

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,D The fifth category is Waste Treatment,-which includes all processing performed on gaseous, l

liquid and solid waste str'eams (generated within the facility), prior to their release from the

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' facility, to enable compliance _with applicable regulations, including the philosophy of."as low as reasonably achievable'* (ALARA).

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Figure 3-1 of the ANFFP Environmental Report shows the general arrangement of the various areas in the Alabama Nuclear Fuel Fabrication Plant (ANFFP) Special Nuclear Materials (SNM)

Building.

O General Nuclear Criticality Safety The nuclear criticality safety of limited moderation (moderation controlled) processing operations t:kes advantage of the fact that, for low enriched uranium fuels [ enriched to less than or equal to 5 weight percent (w/o) U 235], a moderator must be present to attain criticality. The de-scribed IDR (dry) conversion process, with its attendant bulk blending and storage systems, rilies on moderation control (as the primary nuclear criticality safety control in low enriched uranium operations) to a somewhat greater extent than previous!y utilized in Westinghouse fuel fabrication facilities; however, these processes and systems have been thoroughly proven in commercial application with some 11 years of safe operation at British Nuclear Fuels Limited (BNFL - Great Britain) and some two years of safe operation at. Franco-Belge de Fabrication da Combustibles (FBFC - France). In tnese processes and systems, equipment geometries and uranium masses can exceed the Maximum Permissible (Limit) Value (MPV) specified in para-graph 5-2 of this license application; thus, the nuclear criticality safety of individual units, and trrays of such units, is primarily based upon rigorous control of moderating materials (includ-ing water, plastics, porosity control and binder materials for ceramic fuel, and lubricating oils).

A systematic analytical approach was used to demonstrate application of the " double contin-gency principle" and to identify administrative and engineered controls required for safe opera-tion with limited moderation systems.

The nuclear criticality safety of geometrically favorable (geometry-controlled) processing op-erations is established using the approved maximum permissible (limit) values for solutions (see figures 15-5, 154, 15-9, 15-12), and/or subcritical limits (for homogeneous and heterogeneous oxides: see TID-7016; Revision 2, (figures 2.13, 2.14, 2.15, 2.16), in simple geometries. Such limits are typically based on published values, and are selected to be nucleartv safe given op-timum light water moderation and full reflection for isolated subcritical units. (Such units are called subcrits.) The spacing of subcrits is then established using approved empirical rules which assure nuclearly safe separation.

The nuclear criticality safety of fuel assemblies and fuel assembly arrays evaluated is bar on sophisticated computer calculations.

Specific evaluations of future situations which are too complex to be subject to licensed con-trols, and future evaluations of new types of processes and equipment, will be submitted for review and approval by the U.S. Nuclear Regulatory Commission (NRC) prior to implementa-tion.

O D-3.c

Water Barrier Moderation Controls - For portions of the ANFFF' SNM Building operated on a basis of nuclear criticality safety control by moderation limitation ("the moderation-controlled area"), it is essential to prevent entry of water or other hydrogenous liquids (and certain solids) into process equipment, vessels, and containers. The following design principles have therefore been followed:

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The moderation-controlled area is protected against the ingress of the external water (or other hydrogenous liquids), under normal and accident conditions, from areas not subject to moderation control. (For example floor leals are such that flooding into the modera-tion control area cannot take place.)

s The building construction (roof and walls) is of high integrity, and surrounds a secondary high integrity " water barrier" enclosing the moderation control area.

Equipment is designed to minimize the use of hydrogenous materials as services; only process feed lines essential to the enclosed operation are permitted to pass through the water barrier, and only analyzed hydrogenous materials and evaluated material containers specificclly authorized by the Criticality Engineering Discipline may be introduced to the moderation control area.

Figure 15-1 is utilized to demonstrate the integrity of the water barrier, under normal and P

accident conditions, in this safety analysis. All possible sources ei moderator to the moderation-controlled nuclear material have been investigated, and controls have been installed which con-form with the double contingency principle.

' All possible sources of moderator to the moderation-controlled area (through service flow, etc.)

are minimized to the extent practicable; then, those remaining are evaluated using established double-contingency principle analysis techniques.

Processing Operation Moderation Controls - The nuclear criticality safety of large units of low-l' enriched uranium, and arrays of such units, based on moderation control where the moderator i

is assumed to be uniformly distributed throughout the fissile material, has been well docu-Q mented in the literature (and has been in general use by the nuclear industry for many years).

b Moderation control can be logically extended to additional applications by establishing suit-able administrative and/or engineered safeguard,, which necessarily involve consideration of I

postulated normal and accident conditions in.which the moderator might not be so uniformly l

distributed. A'nearly infinite number of combinations of uranium mass and density, moderator g

(v) mass and density, and uranium-moderator system geometry can be postulated for such applica-tions.-(Some configurations would require more moderator than a uniform distribution; for example, as a result of tunneling, layering, or concentration on vessel walls. Likewise, con-figurations in which the moderator is concentrated in.a reactive geometry could require much in-D-3.7

less moderator than a uniform distribution, with the limiting case being the optimum fuel-moderator ratio in a fully reflected spherical geometry.) Therefore, to develop boundary con-ditions for the safety analyses, ti.e most restrictive conditions possible were first determined, using UO -water data from ARH-600 (reference 3 in paragraph 5-2.4 of this License Applica-2 tion), and plotted in figure 15-2. The ARH 600 data are found to be in substantial agreement with the data in the UKAEA Handbook (reference 4 in paragraph 5-2.4) at concentrations up through the minimum critical mass (but is shown to result in somewhat smaller critical volumes at higher uranium concentrations); the ARH-600 data are also found to be in substantial agree-ment with UO -H O and UO F -H O data in the Handbook der Kritikalit t' (reference 13 in 2 2 22 2 paragraph 5 2.4). These data are also plotted in figure 15 2 (which graphically illustrates the wide range of critical conditions which can exist).

Figure 15-3 is an envelope plot of the published critical data which show the relationship of uranium mass, water volume, and uranium-water system volume (as a function of s/ stem uranium density) for 5 w/o enriched UO2 n a spherical geometry with a full water reflector.

i The minimum critical mass of uranium is shown to occur in the range attained by UO F2 2 and UO2 p wders in the evaluated IDR conversion process, and the minimum critical water and uranium water system volume are within the range attained by UO2 p wders in the evaluated bulk handling process. (The broad minimum water volume is relatively constant for a wide range of UO2 masses and densities.)

The previously described plotted data, divided by a 1.1 safety factor, supports the following maximum subcritical moderation levels in homogenous UO, at an enrichment of 5 w/o U-235:

2 Maximum permitted water volume in aqueous reflected UO liters:

19 2

Maximum permitte J weight percent water in an infinite UO2 system-weight percent:

1.0 Maximum permitted hydrogen to uranium atomic ratio in an infinite UO2 system-H/U:

0.5 The minimum quantity of water required for criticality increases as uranium enrichment falls below 5 w/o U-235, and as less than full reflectio. is provided. This is shown in figure 15-4.

Therefore, con'rols on water equivalent volumes based on 5 w/o enrichment and full reflection are shown to be conservative at lower enrichments and/or less than full reflection.

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Safety Factors and Maximum Permissible (Limit) Values for Moderated Solution Systems -

Maximum Permissible (Limit) Values (MPV) for isolated solution units have been established using published data. The data were converted to minimum critical values in standard units (as required), and plotted on graphs. Such minimum critical values assume optimum water modera-

- tion and reflection.

MPV's were developed using the safety factors listed, where:

, Minimum Critical Value.

p Safety Factor O

Parameter Safety Factor Mass (Administrative Control) 2.3 Mass (Engineered Control) 1.3 Concentration -

2.0 Volume '

1.3 Slab Thickness 1.2

' Cylinder Diameter 1.1 The values established-for aq'ueous solutions are applicable to uranyl fluoride (UO F ) systems, 22 to uranyl nitiate -[UO (NO )2] systems, and to' moderated systems which have a theoretical 2

3 3

uraniu'm density less than 3.2 g U/cm.'When aqueous solutions are converted to oxide powders,-

the TFD-7016-(Revision 2) subcritical (homogeneous) limits for UO ;are applicable to the.op-2 eration. (The UO2 values apply to all uranium oxides.)

Minimum critical values and MPV's,for solutions, are summarized in figures 15-5, 6, 9,.

and 15-12.-

Cylinder data are generally ~ presented as for circular cylinders, but are equally applicable (on

. an ' equal cross sectional area basis) to cylinders whose cross sections have other shapes.

1 The. diameters of-intersecting pipes are reduced 3 n accordance with procedures in TID-7016, i

Revision'2, except.that the permissible pipe diameters are determined by multiplying the ap-plicable MPV or O.3 fraction critical vaiue"lbyLthe following factors:

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intersection Factor Ells 0.92 Tees 0.84 Crosses or Wyes 0.76 These factors were derived from the Tf D-7016 values for full reflection.

Nuclear Interaction Evaluation of Moderated Systems. - All subcrits must meet the MPV's and/or subcriticallimits,and there must be a separation of at least one foot (edge to-edge) between any two subcrits; nuclear interaction evaluations are conducted for all subcrits which are not nuclearly isolated.

Interaction between subcrits which are not isolated is evaluated using one or more of three methods: (1) surface density, (2) solid angle, or (3) KENO-IV Monte Carlo computer code.

The surface density method requires that each subcrit be limited to the nuclear equivalent of a sphere having less than (or equal to) 0.3 times the volume of an unreflected critical sphere of the same material ("<0.3 fraction critical"). Values are established for each control using equal buckling conversions as follows:

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(R +Acj (H+2A R +A s sj c

c for a cylinder of finite length; Where:

R = the radius of a sphere T = finite slab inickness L = finite slab length W = finite slab width R = finite cylinder radius c

H = finite cylinder height A = 2.1 cm for unreflected spheres s

A = 2.8 cm for unreflected slabs t

A = 2.5 cm for unreflected cylinders.

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(N The calculated 0.3 fraction critical (0.3 F.C.) values for geometry-controlled subcrits are shown in Figures 15-6 through 15-14. The value for mass is sufficiently low (0.435 of the reflected critical mass) that it is much less than 0.3 of the unreflected critical mass assuming a 3.9 cm i

reflector savings for volumes with concentrations above safe concentration limits.

The subcrits are centered in a spacei determined by the ratio T /T, where T is the " smeared" s c s

thickness (the thickness of the slab which would be formed if all the Special Nuclear Material (SNM) permitted in the subcrit were to flow out over its assigned space), and T is the mini-c mum critical reflected infinite slab thickness for the SNM being considered. For a given suberit the fractinn critical value will be limited to < 0.3 and the ratio T /T will be limited to < 0.25.

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For geometry-controlled subcrits, the respective fraction critical values are determined on an equal volume basis. That is, the assigned surface density area, times the appropriate T /T s c

. (<0.25), times the minimum critical slab thickness (from Figures 15-12,-13, or -14), is greater than (or equal to) the volume of the suberit. Mass-controlled subcrits are spaced such that the

- average surface density in the ' assigned area is < 50% of the minimum critical surface density for UO2 (homogeneous or heterogeneous) or UO F2 2 reflected slabs. This is plotted on Figure 15-15.

If array interaction is evaluated using the solid angle method, the spacing criteria specified in f

TID 7016, Revision 2,- will apply The maximum allowable subcrit multiplication factor (k,gg)

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will be limited to 0.8.

When array interaction is evaluated using KENO IV Monte Carlo computer' codes, only validated codes are utilized. In each case, the maximum allowable value 'of k gg + 2a will be limited to e

< 0.95.

' Moderation-controlled units (H/U < 0.3), and units poisoned with neutron absorbers,'are specifically excluded from these interaction considerations, provided that they are not located in an area designated by the surface density method for other subcrits, and provided that they

. are *ccated 'at least one foot from other SNM. Transfer piping <2 inches in diameter, and

. safe diameter vent lines, are also excluded from interaction calculations.

V Neutron Absorbers - Fixed neutron absorbers (" poisons'.') are used only in nonliquid systems, L

'in situations where the absorber-can be physically protected from abrasive action by the con-tained special nuclear material. Nuclear criticality safety of such poisoned systems is always verified by validated computer calculations. Typical of such use would be the addition of

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stainless-steel-clad (axial) boron. carbide rods to conversion kiln check hoppers.-

Borosilicate glass Raschig rings are used only in liquid systems, in accordance with Regulatory Guide 3.1; "Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber ~in Solutions of n

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. LEG END:

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7 8 9 10 U 235 ENRICHMENT, WElGHT PERCENT Figure 15-8. Volume Data Heterogeneous UO -Water 2

D-3.19

.. ~.. -

17696-9 9

CRITICAL UNREFLECTED CRITICAL HEFLECTED l

0.3 F.C.

)

11.4 10.4 e

g my 9.2

~

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y 1.1 S.F.

3 8

g Q

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o LEGEND:

O T D 7028 6 K 1629 l

l l

l l

l l

g 1

2 3

4 5

6 7 8 9 10 l

l

~ U 235 ENRICHMENT, WElGHT PERCENT Figure 15-9. Cylinder Data for Solutions O

D-3.20 l

-, -. -,.. -..,,...,. -. -, ~, -

17696-10 LEGEND:

O AHSB HANDBOOK 6 DP 1014 V K 1629 0 W LEOPARD CALCULATIONS O

CRITICAL UNREFLECTED CRITICAL REFLECTED 5z

)

g 0.3 F.C.

r 10 9.0 0.4 F.C.

~

1.1 S. F.

l O a

g 0

1 2

3 4

5 6

7 8 910 U 235 ENRICHMENT, WElGHT PERCENT Figure 15-10. Cylinder Data Homogeneous UO -Water 2

O D-3.21

17C36-11 LEGEND:

O AHSB HANDBOOK O DP 1014 6 W LEOPARD CALCULATIONS CRITICAL UNREFLECTED CRITICAL s

REFLECTED

~

0.3 F.C.

y 0.4 F.C.

8.0 e

i w

m 1.1 S. F.

lE Q

5 O

2 N

o I

h 9

i i

i 1

3 2

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5 6

7 8 9 10 U.235 EN 'lCHMENT, WElGHT PERCENT Figure 15-11. Cylinder data Heterogeneous UO -Water 2

D-3.22

17696-12 LEGEND:

O TID 7028 i

O

~

i CRITICAL l

UNRE FLECTED lO 10 E

0.3 F.C.

ba 0.4 F.C.

h

.3 1.2 S. F.

-6 5

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5.2 CRITICAL l

a REFLECTED i

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~

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]

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=

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g., SMEARED" j

i l

i I

1 2

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5 6

7 8 9 10 U-235 ENRICHMENT, WEIGHT PERCENT Figure 15-12. Stab Data for Solutions O

D-3.23

17696-13 0

LEGEND:

C AHSB HANDBOOK A DP 1014 O W LEOPARD CALCULATIONS CRITICAL O

UNRE FLECTED CRITICAL REFLECTED S

I 10 z

[-

0.3 F.C.

d 0.4 F.C.

z U

1.2 s

S. r.

r 4.3 i

l l

50% "SM EARED" 20% "SM EAREO" r }

30% "SM EARED" 1

I I

I I

I I I l

1 2

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5 6 78910 U-235 EN RICHMENT, WElGHT PERCENT 1-Figure 15-13. Slab Data for Homogeneous UO - Water 2

O D-3.24 l

l l

17696-14 LEGEND:

O AHSB Handbook O DP 1014 6 W LEOPARD CALCULATIONS 10 CRITICAL UNREFLECTED 0.4 F.C.

E 1.2 S. F.

CRITICAL I

REFLECTED 0.3 F.C.

_5 l

I 3.5 i

\\

O.'

j vs 50%

"SM E A R E D" 20% "SM E AR ED"

$ MEARED" I

l I

I 1.0 1

2 3

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6 7 8 9 10 U-235 ENRICHMENT, WEIGHT PERCENT

,,,,,,5 14. S,.e D.,.,., N.,,,e _, U D, _

D-3.25

17696-15 O

100 LEGEND:

O DP 1014 6 TID 7028 - UO F 22 MINIMUM CRITICAL REFLECTED N

c 3

E

(

a 2

uJ Q

5 x

D O

m E<

10 lE 50% "SME ARED" 5.15 O

1 I

I I

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I 1

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6 7 8 9 10 U-235 ENRICHMENT, WElGHT PERCENT l

Figure 15-15. Data for Mass Surface Density 1

O f

l D 3.26 l

I....

d Fissile Material". Raschig rings may be used to provide contingent (secondary) nuclear critical-V ity safety assurance in vessels of unfavorable geometry, where normally uranium concentrations 235U/ liter are maintained less than the maximum permissible solution concentrations (5.0 g for aqueous solutions) as the primary nuclear criticality safety control. Typical of such uses would be the addition.of Raschig rings into the process quarantine tanks which are sampled prior to release for waste treatment and addition of Raschig rings to effluent scrubber system L

reservoirs.

The use of any other nuclear poisons, or any use of poisons for primary criticality control, re-quires a specific license authorization.

p-h Fuel Rods and Assemblies - A safe slab thickness for close-packed, clad, L'0 f

2 uel rods has been calculated by Westinghouse Nuclear Engineering. These calculations were conservatively based on a square lattice. K= for a triangular lattice was less than 0.95 for Zircaloy-clad rods containing uranium enriched 45.0 w/o. The maximum permissible slab thickness (0.84 of the minimum water moderated and reflected critical slab thickness) is 15 inches for UO2 r ds with a U-235 enrichment <5.0 w/o. The unreflected critical slab thickness is: (Maximum Permissible Thickness /0.84) + (2 X 3.9/2.54). This equals 20.9 inches for a U 235 enrichment <5.0 w/o.

The "0.3 fraction critical" limit for use with the surface density spacing of moderated subcrits is conservatively determined using 0.3 times this critical unreflected slab thickness, or 6.3 inches, for <5.0 w/o U-235 enrichment. Similarly, the 25% of reflected critical thickness for surface density spacing is 4.6 inches for <5.0 w/o U-235 enrichment. All maximum permis-sible fuel rod slabs are. separated by at least one foot. The arrays, which are dry under normal conditions (and are controlled so as not to retain water), are not considered in interaction evaluations.

Fuel assembly k,ff numbers are calculated assuming maximum moderation and full reflection by light water, since assemblies have consistently been designed to be undermoderated when flooded. Such calculations would assume optimum moderation vehen (and if) such moderation

is considered physically possible.-The maximum fuel assembly keff s 0.95. Fuel assemblies i

are also separated by at least one foot, and are also controlled so as not to retain water.

[

lnteraction with other fuel assemblies is, therefore, not considered.

m General Radiological Safety -

The radiological safety of processing operations evaluated takes advantage of the 11 years of safe operating experience at the Westinghouse nuclear fuel fabrication plant in Columbia, South

' (j Carolina. The Alabama Nuclear Fuel Fabrication Plant is designed to enhance this 3xperience through application of improved technology and prudent radiological safety practico developed and recognized by health physics professionals throughout the international nuclear community.

These practices are embodied in the Regulatory Compliance Manual, the ALARA Manual, and OL)

D 3.27

the Emergency Plan (submitted in support of this license application); and, these practices are r:flected in this safety analysis.

O General Occupational Safety and Health The occupational safety and health aspects of processing operations evaluated is based upon the Westinghouse commitment to provide safe and healthful working conditions and reduce occupational injuries and illnesses to the lowest practicable level (the Columbia fuel fabrication plant accumulated more than 8 million manhours worked without a lost time occupational in-jury or illness), and, upon the Westinghouse policy of compliance by all locations with both the spirit and intent of federal, state and local legislation and governmental regulations provid-ing for occupational health and safety.

15-2.1 Conversion The Integrated Dry Route (IDR) dry conversion process is utilized to convert solid uranium hexafluoride (UF ) to uranium dioxide (UO ) p wder. Uranium hexafluoride is received for 6

2 input to the process, which involves UF6 vaporization, gas phase hydrolysis, and gas-solid phase reduction to produce UO. The UO2 p wder from the dry conversion process (or out-2 side sources) is transferred to a powder processing area where it is blended with additives, in-ciuding uranium oxide (U 0 ) recycled from scrap recovery processes, in preparation for (ship-38 ment or) subsequent fabrication operations.

Cylinders - Uranium hexafluoride is received, from a 15-2.1.1 Receipt and Handling of UF6 diffusion enrichms.nt facility supplier, within Model 30A or 308 cylinders in NCR and DOT authorized packagings. (These cylinders normally arrive by truck, in their protective shipping containers.)

Upon arrival, the packages are removed from the carrier (using a lift truck) vehicle. The cylin-6 are then removed from the outer packages, are individually weighed, and are then 1ers of UF Storage Area specifically transferred via lif t truck (into the SNM Building or) to a secure UF6 Storage Area, the cylinders are stored horizon-designed for such cylinder storage. In the UF6 tally on wocden chocks (or in protective overpacks).

Based on need, the UF6 cylinders are transferred (using a lift truck) into the vaporization staging area of the SNM Building where they are stored horizontally in fixed steel racks prior to subsequent processing. (Inside transfers are made by overhead crane.)

i 6 Storage Area 6 nto the process, the cylinders are returned to the UF After removal of UF to await return to a supplier.

O D-3.28

a Nuclear Criticality Safety The maximum U-235 enrichment shipped in Model 30A and 308 cylinders is 5 weight percent (w/o).

6 n 30-inch-diameter cylinders based on moderation i

The nuclear criticality safety aspects of UF O

control is documented in Reports K-1663 and K-1920 (references 8 and 12 in paragraph 5-2.4 of this license application). Uranium hexafluoride in such a standard cylinder constitutes a moderation-controlled subcrit (when outside the conversion system). These cylinders are stored in a one-high planar array with a minimum of one foot edge-to-edge cylinder separation, thus they require no interaction considerations. (in UF6 Storage Area, the chocks or overpacks A

maintain the required spacing; in the vaporization staging area, the fixed steel racks maintain the required spacing.)

Storage Area, being within the Special Nuclear Material Building plant area, is The UF6 equipped with a nuclear criticality detection (and alarm) system meeting applicable criteria described in 10CFR70.24 and applicable license specifications.

a Radiological Safety Upon arrival at the site, the packages are surveyed (direct radiation and radioactivity), by the Regulatory Compliance Component, before being unloaded from the carrier.

Low enriched uranium in sealed cylinders requires only routine radiological safety precautions.

At no time is a cylinder more than 20 feet above an unyielding surface; thus, no damage j

should be expected even if a drop occurred. (During published tests of UF6 cylinders for qualification, it has required a 30 foot drop to develop even a hairline crack :n such a cylin-der; and 'even if a hairline crack should occur, it would not cause a major leakage of the 0

6 s a solid at ambient temperature (it sublimes at 132 F), and therefore would i

UF. The UF j

6 I

seep from a crack very slowly. Further, UF6 reacts with atmospheric humidity to form uranyl fluoride (UO F ), a nonvolatile solid; thus, a slow leak would tend to be self sealing.] Only 22 uranium hexafluoride cylinders that have been processed and evacuated, and have been so l

l O

identified, may be moved without protective valve covers in piace.

l l

-Cylinders are resurveyed (direct radiation and radioactivity) after removal of the UF Pfi f 6

to return to the Storage Area.

.m Occupational Safety and Health l

,a y

Combustible materials are kept to a minimum in UF ' cylinder storage and staging areas.

l V' 6

f.

. All precautions associated with moving high-pressure cylinders are observed.

i

( v D-3.29 I

Transfer and handling of UF6 cylinders is in accord with applicable lifting device safety piac-tices embodied in Site Regulatory Compliance Manual Procedure RCM-501.

15 21.2 UF6 Vaporization - Full UF6 cylinders are transferred from the UF6 Storage Area to the receiving area of the SNM Building by lif t truck. In the SNM Building, cylinder identi-fication and cylinder we:9nt are inspected by the Quality Control Component and are recorded; thus, the cylinders are " released" for processing.

Quality Control released cylinders are transferred via an overhead crane, and installed in a vaporizer chest. The UF6 cylinder (connected to its process header by flexible copper tubing) is heated to a temperature of approximately 180 F, in the vaporver chest, by circulating hot water sprays. When a cylinder has reached process temperature and pressure, it is ready for opening of the connection to the UF6 kiln supply header (by use of a mechanical thru wall activated valve). After a cylinder (which is supplying the conversion system) is sufficiently depleted of UF6 so as to no longer maintain a supply pressure above 5 psig, it is remotely disconnected from the supply line and valved into a cold trap evacuation system foi removal 6 o an acceptable final " heel" of less than 25 pounds.

of residual UF t

Heel removal is accomplished by evacuating the cylinder with a vacuum pump through an ex-haust train consisting of: a cold-trap system with self contained refrigeration (-65 F) to con-6 vapor, and two final chemical (Al O ) traps (in series) for the capture of any dense UF 23 final traces of UF. Upon completion of the evacuation procedure, the cylinder is removed 6

from the vaporizer (using the overhead crane) and transferred to the cylinder scales.

The cylinder is weighed to assure that the residual heel is less than (or equal to) 25 pounds; the cylinder pressure is measured to assure 18 to 22 inches (Hg) of vacuum. (The cylinder identification, cylinder weight and cylinder vacuum are inspected by the Quality Control Com-ponent and are recorded.) If the UF6 residual heel weight is in excess of 25 pounds, the cylinder is returned for heel removal.

Storage Area to await After Quality Control inspection, the cylinder is transferred to the UF6 r: turn to a U.S. Government owned diffusion enrichment facility for reloa6ng.

O Nuclear Criticality Safety During heating and cooling operations, the UF6 cylinders are subject to water sprays only; at no time are the cylinders immersed in water.

The maximum pressure during vaporization is about 43 psia. Since the cylinders are con-structed to a pressure venel specification, and are subject to regular test and inspection, it is not considered likely that a major failure would occur under this operating pressure. However, on the conservative hypothesis that a minor failure might occur, the low holdup of water in O

D-3.30

_~

__~

/

Ib-[

the safe slab geometry liquid reservoir (with overflow to the vaporizer safe sump) provides con-1 tingent nuclear criticality control.

.e During cylinder cooling, a negative (condensation) pressure will develop in the vessel, but the maximum differential pressure would be less than 15 psia. This might lead to the entry of some water into the cylinder in the event of a failure, but a significant influx is unlikely for the following reasons: even on a most conservative assumption of failure it is likely that this would occur only at a weld joint, or in connective apparatus, and hence would be small in total area; spray cooling is used to reduce the amount of water likely to be drawn in; a posi-tive pressure is likely to develop due to prcduction of HF gas from the reaction between UF6 and water; and, in the event of an unexpected increase of condensation pressure, provision is made for introduction of nitrogen (up to atmospheric pressure).

The lower portion of the vaporizer chest is designed to provide a liquid rc.ervoir with over-flow to a vaporizer safe sump. (A conductivity cell immediately downstream of the reservoir 6 eakage so that appropriate actions can overflow weir will provide early warning of any UF l

be taken to promptly terminate the leakage; a high conductivity alarm automatically shuts off hot water spray to the system and UF6 flow from the system.) The vaporizer safe sump is a closed floor unit designed to contain uranium in liquids (as a result of any leakage from UF6 vessels, vaporizers, and/or connective piping) in a safe slab geometry of less than 5.2 inches depth (also maintained by overflow weirs). Further, the vaporizer safe sump is arranged to v

also act as an emergency vent duct to carry uranium in gases as a result of any leakage from UF6 vessels, vaporizer, and/or connective piping to the HF vent scrubber; and, to provide the necessary holdup for analysis of any liquids (for uranium content) prior to potential transfer to unfavorable geometries.

All other process vessels and equipment (i.e., except for the cylinders and vaporizers previously evaluated) in the UF ' vaporization area, normally containing uranium or having the potential 6

to c.,ntain uranium as a result of accident conditions, are designed safe by geometry for 5 w/o U 235 materials under full reflection and optimum moderation.

p_

~ lnteractions under normal operating conditions will be negligible due to the low reactivity (con-Lij trolled moderation) of the material and the separation space between vessels. It is not con-sidered likely that major failure of two adjacent units could occur simultaneously, a

Radiological Safety (6_p-Low-enriched uranium. in the closed. vaporization system requires only routine radiological

. safety precautions. All connections are pressure-tested to assure they are leaktight prior to in-troduction of UF._ Each. vaporizer chest is equipped with a~ removable cover, which is capable 6

of sealing the chest at a pressure of 12 inches of water, and a through-wah (manual) remote valve operator for the UF6 cylinder discharge valve.

V l

1 D 3.31 -

,,9

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.w-,

cy,<-

e,

,,er

-,=,

,r v s r*v-t' d--w

- --= * * --

Emergency UF6 cylinder cooling is provided by a spray system connected to an external emergency cooling head tank.

A vaporizer chest is connected to its liquid reservoir through a series of downcomers which extend below the surface of liquid in the reservoir, thus assuring that any gases or vapors from the vaporiter chest would be bubbled through the liquid reservoir to the gas space of the va-porizer safe sump (to assure hydrolysis of any inadvertently released UF ), pri r to venting 6

through the HF Vent Scrubber system.

The area of the " pigtail" (flexibb copper tub'ng) connection in a vaporizer chest is suitably baffled and connected into the HF vent suubber system (to provide and maintain an inlet face velocity of at least 150 linear fect per minute of airflow into the vaporizer opening when the pigtail is disconnected). The process header valves are interlocked such that, when a UF6 cylinder is removed and its line block valve is closed, the relevant kiln feed valve and cold trap bypass valve will also be closed (thus preventing backflow of UF6 ut of the syNem).

Any leaks in the UF6 confinement system outside the cylinder and/or vaporizer chest would be controlled by valving off the applicable line. Further, all UF6 v p r ( r liquid) bearing lines and vessels (including those of the cold trap evacuation system) are enclosed within (pipechases or other) conteinment, maintained at a negative pressure with respect to the op-erating area, and vented through use HF vent scrubber system.

The UF6 vaporization room is designed to enable physical isolation from the remainder of the conversion process. The UF6 vaporizer room normal (HEPA) ventilation system inlet (and selected UF6 confinement space) is continuously air monitored (CAM) for radioactivity, and the monitor alarm is interlocked to divert ventilation (and selected UF6 confinement vents) to the HF vent scrubber system in the unlikely event of an accident leading to UF6 release to (selected UF6 confinement space and/or) room air.

c Occupational Safety and Health Operations personnel are provided with rubber gloves and full-face shields as additional pro-tective equipment for making and breaking pigtail connections.

The vaporizer chest is connected to the liquid ressrvoir by a series of downcomers which provide a pressure release mechanisrn for the chest interior.

6 r m the f

6 o UO2 - Reactants, consisting of gaseous UF 15-2.1.3 Dry Conversion of UF t

vaporization process and superheated steam, are continuously introduced to the (feed end) dis-entrainment chamber of a conversion kiln. Here, the UF6 vapor and superheated steam undergo dry hydrolysis to form uranyl fluoride (UO F ) p wder and hydrogen fluoride (HF) gas. The 22 UO F2 2 powder falls to, and is continuously removed from, the bottom of the chamber and O

D-3.32

'J slowly tumbles down a slightly slanted (and rotating) kiln barrel. Hydrogen gas and super-heated steam are continuously introduced at the (discharge) product end of the kiln. In the U

kiln barrel the powder is further contacted with a countercurrent flow of this hydrogen and steam, to strip the remaining fluoride and reduce the uranium to uranium dioxide (UO )

2 fm powder.

('")

Off-gases (hydrogen, hydrogen fluoride, nitrogen, and steam) are continuously removed from the kiln system through process filters at the top of the disentrainment chamber (for temporary retention and subsequent.titrogen blowback of uranium powder). A conversion kiln is main-tained under a controlled positive pressure (0.3 to 0.9 inches Hg) by means of the ventilation 7

3 V

off-gas system (the higher pressure being reached during process filter blowback). Nitrogen-purged lip seals at each end of the kiln barrel (seal pressure > kiln pressure} enable rotation without loss of confinement. The conversion kiln system is operated with zone-controlled temperature conditic: s typically hydrolysis (disentrainment chamber) at 130 to 300 C, reduc-tion (kiln barrel) at 630 to 770 C, and (process) filters at 200 C.

Uranium dioxide powder is removed continuously from the (low) discharge end of a kiln, into one of a pair of UO2 check hoppers. Provision is made for continuously purging the UO2 check hoppers with dry nitrogen during filling and analytical hold time. Product entering a UO2 check hopper is continuously sampled by a proportional sampler. The samples are analyzed for

/

')

acceptable quality with respect to both fiuoride content and moisture content. Acceptable

'"/

product is disenarged to bulk powder ntainers for transfer to powder preparation processes;

\\

powder in a check hopper found to be unacceptable with respect to fluoride and/or moisture content is discharged to special containers for transfer to powder rework processes, Nuclear Criticality Safety a

1.

Moderation Control Contingencies The IDR conversion operations rely priacipally on moderation control for nuclear criticaiity safety. Operations are designed such that multiple contingencies (accidents, failures, or events) must occur before conditions for a criticality accident are possible. The systems are evaluated 7

by postulating contingencies which could result in sufficient moderation getting into sufficient enriched uranium in an unsafe geometrical configuration, and by addressing the controls which prevent these events from occurring. (At least double-contingency criteria are satisfied for each scenario.)

[

Scenario: Sufficient Liquid Water Enters Diwntrainment Chamber Through Steam Line For liquid water to get in through the steam supply line, the boiler system must fail (result-ing in the boiler being solid with water), and the water trap must fail, simultaneously. How-ever, if these failures in the steam supply system occurred, the steam temperaturehteam feed

(

D-3.33

[

interlock / alarm system would shut off the steam valves, thus stopping flow to the chamber.

Even if this interlock / alarm system also failed, the chamber temperature interlock / alarm system would shut off the steam valves, also stopping flow to the chamber. (Although the steam valves are common to both of these interlock / alarm systems, both an in line process control st:am valve and a safety shut-of f steam valve - located in series - must fail to close for the interlock / alarm system to fail.)

[lt is not considered credible for sufficient liquid water to enter the charnber through the uranium hexafluoride lire (UF6 " dry specification).]

Scenario: Sufficient Stram Enters Disentrainment Chamber Under Normal System Pressure Condensation Conditions Normal steam input to the chamber could condense if there is a loss of heat input to the chamber; or, wet steam input could condense because of inadequate heat input to the chamber.

To get we. steam into the chamber, there must be a failure in the steam supply system, the superheater system must fait, 2nd the steam temperature / steam feed interlock / alarm system must fail to shut off the steam supply. If there is a loss of heat input, or inadequate heat input, the chamber temperature interlock / alarm system wc,ald shut off the steam supply valves.

Even if these interlock / alarm systems fail, such that steam input continues under condensation conditions, an administrative failure (alarm response) to take corrective action would also be required before sufficient steam could enter the chamber.

Scenario: Sufficient Steam Enters Disentrainment Chamber Under Abnormal System Pressure Condensation Conditions.

At very high (abnormal) pressures (approximately 150 inches Hgi steam condensation could occur at temperatures within the normal operating r;nge (>130 C) of the chamber. To ob-0 tain such pressures, all of the sequenced blowback filters must plug, the steam regulator must fail, and the seals in the chamber must hold the high pressure. Then, for sufficient steam input er. der these failure conditions, the pressure interlock / alarm system, which shuts cff the steam supply, must also fail.

Scenario: Failure Initiated by Disentrainment Clw.nber Power Interruption The loss of both normal and emergency power could result in the disentrainment char,icer cooling to the extent that steam condensation would occur. To get sufficient steam. input under this condition, however, the (reverse acting fail-safe) process valve and the (reverse acting fail-safe) safety valve in the steam supply line must both fail to close when both of tue power systems fail.

\\

O D-3.34

l q

Scenario: Sufficient Water Enters Kiln Through Steam Line

'U The initial failures required for this event, including the steam temperature / steam feed interlock /

alarm system failure, are the same as those for water entering the disentrainment chamber through a steam line. The kiln barrel also has a temperature interlock / alarm system set to shut s.

off the steam line feed into the kiln. (The process steam control valve and the safety shut off valve are common to the interlocks / alarms on the kiln, as well as to the disentrainment cham-ber).

[lt is not considered credible for sufficient liquid water to enter the barrel through the hydro-2 s not introduced until the kiln is at operating tem-gen line (H2 " dry" specification and H i

/

perature) or through the nitrogen line (cryogenic supply).]

Scenario: Sufficient Steam Enters Kiln Under Condensation Conditions As with the disentrainment chamber, normal steam input to the kiln could condense if there is a loss of heat input to the barrel; or, wet steam input could condense because of inadequate heat input to the kiln. To get wet steam input, there must be a failure in the steam supply system, the superheater system must fail, and the steam temperature / steam feed interlock /

' alarm system must fail to shut off the steam supply. If there is a loss of heat input, or in-adequate heat input, the kiln temperature interlock /alsrm system would shut off the steam supply valves. Even if these interlock / alarm systems fail, such that steam input continues under

-V-condensation conditions, an administrative failure (alarm response) to take corrective action would also be required before sufficient steam could enter the kiln. (The kiln operating tem-perature is much higher than that of the disentrainment chamber; therefore, abnormal pres-sures at which steam could condense at normal operating temperatures are not possible.)

Scenario: Failure Initiated by Interruption of Power to Kiln As described in the disentrainment chamber, if normal and emergency power to the kiln are both lost, a fail-safe system closes both the process and safety steam supply valves.

Scenario: Sufficient Moderator in Moderation-Controlled (Modcon) Container

[i i

(,/

!The kiln discharges into one of two check hoppers where the UO2 s quarantined for moisture analysis. The check hoppers have permanent axial nuclear poison rods so that moderation con-2 s discharged into moderation-trol is not essential. Following moisture analysis, the UO i

controlled (modcon) containers if it is " dry", or into geometrically favorable (geocon) containers O

if the moisture content exceeds the " dry" limit. For sufficient moderator to get into a modcon V

container (with sufficient UO ), it must be introduced into the container prior to filling with 2

UO, the UO2 must contain the moderator, or the moderator must be introduced after the 2

container is filled.

H

~

D-3.35

Administrative procedures restrict the handling of modcon containers within carefully segregated

(" water barrier") moderation control areas where moderators are strictly limited; thus, these prccedures must fail before moderator could get into an empty container. Modcon containers tre closed and sealed while empty and not in use. Further, inspection procedures require visual checking of each container prior to hookup for filling; thus, these procedures alsc must fail before UO2 could be put in a container holding moderator.

fr m a check hopper into a modcon container, there must first To discharge moderated UO2 2 n the check hopper to become mod-i have been some process upset which caused the UO erated. Further, the normal discharge system is interlocked / alarmed such that both a powder-moisture monitor and a nitrogen purge gas humidity monitor must fail (or, their interlock /

alarm sy;tems must fail) before discharge of moderated material into a modcon container is i

possible. In addition, the administrative procedure requiring sample analysis of the UO2na 2 s " dry" (or the sample ana!ysis itself) must fail before i

check hopper to verify that the UO moderated material could be discharged into a modcon container.

2 nto a favorable geometry A manual override system for discharging reject (moderated) UO i

(geocon) container is interlocked / alarmed to operate only when a geocon container is in place.

Discharging moderated material into a modcon container through this system would require both the administrative failure of placing a modcon container in place for discharge of the rrjected UO2 and a failure of the interlock / alarm system designed to prevent discharge into the wrong container.

To get moderator into a filled modcon container, the integrity of the container must be breached, and the administrative procedure restricting moderating materials from moderation control areas must fail.

Scenario: Spills From Process Equipment or Containers Three possibilities for spilling sufficient enriched uranium in the moderation cuatrol area are:

a gross containment integrity failure, failure of discharge or feeding mechanism controls by mechanical failure or administrative procedure and interlock connection failure, or administra-tive procedure and instrument failure leading to overfill and subsequent spillage.

The only ways for having sufficient moderator in the moderation control area are a failure of the moderation control barrier (such that moderator could enter the area) or failure of the administrative procedures controlling moderators in the area.

Both these events would have to occur within a short time of each other for the uranium and moderator to get mixed in an unsafe geometry. (if these events did not occur at the same time, the administrative procedures requiring prompt corrective actions would also have to fail.)

O D-3.36

O 2.

Safety Interlocks b

Although steam is introduced into the IDR conversion process, and HF is produced in the reaction, system temperatures are maintained at levels which ensure that these hydrogenous compounds remain in the vapor phase and hence do not act as significant moderators.

N Nuclear criticality safety control has therefore been principally directed to the prevention of significant moderation, because although certain parts of the process equipment are geometrically controlled, technical process parameters have in some cases required sizes in excess of those required for geometry control.

Assurance that significant moderation cannot occur in such process equipment is there.

fore based on the detailed interlock / alarm system _ incorporated in the design to detect conditions which could lead to the presence of l$ydrogenous moderators in the liquid phase and.to automatically. apply corrective actions well before potentially hazardous con-

.ditions could develop. All temperature safety interlock / alarm circuits have redundant thermocouples. (If a single thermocouple fails, an audible / visual alarm is actuated; if both thermocouples fail, the controlled system will automatically shut down.) All safety interlock /aiarm shut off elements are designed to provide a redundant control function independent of the process control instrumentation; that is, separate block valves, in series, are utilized for process control and for safety shut off.

- Further, rigorous' attention is applied to excluding moderators from the secondary water barrier (the room) and-from modcon containers. As a contingent to this effort, additional-interlocks / alarms are installed to prevent inadvertent discharge and/or spills of moderation-controlled powder from the primary water barrier (process _ vessels).

Disentrainment Chamber Interlocks -- The.following' interlocks / alarms in the disentrain-ment chamber process equipment will automatically shut off uranium and steam input to

~

the system upon detection of the indicated fault:

1.

Loss of electrical continuity in (jet) steam superheater circuits 1

~

2.

Low (jet) steam temperature Y

3.

' Loss of electrical continuity in chamber heatsr circuits 4.

Low chamber temperature

'5.

High chamber temperature i

' Kiln Garrel Interlocks The following interlocks / alarms in the kiln barrel process equip-rN

(

[

ment will automatically shut off uranium and steam input to the system upon detection f'

of the indicated fault:

i.

f),v(

[

D-3.37 -

1.

Loss of electrical continuity in barrel probe steam superheater circuits; 2.

Low (barrel prot,e) steam temperature; 3.

Loss of electrical continuity in barrel heater circuits; 4.

Low barrel temperature.

Check Hopper Discharge Interlocks - The following interlocks / alarm in the check hopper process equipment will automatically prevent discharge (and/or spills) of powder from the geometry / poison controlled check hoppers to modcon containers (or the room area)

'upon detection of the indicated fault:

1.

Moisture in powder (including two interlocks / alarms in the moisture probe pro-tection system, loss of electrical continuity in the moisture analyzer circuit and loss of probe cleaning nitrogen flow).

2.

Moisture in hopper purge gas (including four interlocks / alarms in the dew point analyzer protection system, low gas flow to the analyzer, low gas flow to the analyzer filter blowback system, failure of the analyzer filter blowback system timer and loss of electrical continuity in the analyzer circuit).

3.

Low hopper purge gas flow 4.

Inadequate discharge container connection 5.

Overfill of discharge container 6.

Attempted use of geocon discharge manual override without proper (geocon) container m place.

3.

Ancillary Controls Because of the substantial impairment of powder flow properties at even very low moisture content, it is unlikely that significant quantities of moderated material could be trans-ferred to the check hoppers. Moderation of powder already within the (unheated) hoppers could only arise from the back-diffusion of the (kiln barrel probe) steam supply via the discharge end of the kiln. This is a most unlikely event since the steam flow would have to overcome the resistance of the main (kiln) hydrogen gas flow as well as the nitiogen purge which is fed through the base of the hopper. However, as an additional contingency, each cylindrical check hopper is fitted with a stainless steel annular insert containing boron carbide, which runs coaxially within the hopper and serves as a nuclear poison. A KENO calculation has demonstrated the nuclear safety of the check hopper arrangement (the hopper axes are 4 feet apart) containing 5 w/o U-235 materials under nominal (1 inch water) reflection and optimum moderation.

O D-3.38

Because of the (previously evaluated) multiplicity of controls on powder formed in the

(

disentrainrr.ent chamber, it is very unlikely that significant quantities of moderated ma-terial could be transferred to the (heated) process blowback filters. Further, it is even more unlikely that moderation of powder already within these filters could occur since this would require a failure of the process (dry nitrogen) blowback system in addition to O

moderation by faulty disentrainment chamber products. However, the filter arrangement (4-by-2 rectangular array, filter axes 22 inches apart on the long array dimension and 27 inches apart on the short array dimension) was modeled, and a KENO calculation has demonstrated the nuclear safety o'f a fully packed filter array containing 5 w/o U-235 A

materiais under optimum moderation.

Q All other process vessels and equipment (i.e., except for the disentrainment chamber, kiln barrel, check hoppers, and modcon containers previously evaluated) in the dry conversion area, normally containing uranium or having the potential to contain uranium as a result of accident conditions, are designed safe by geometry for 5 w/o U-235 materials under full reflection and optimum moderation.

It is not considered possible that fault conditions could arise to such an extent as to cause interaction between materials, either with respect to adjacent units within a single kiln system or between adjacent kiln systems.

i a

Radiological Safety v

Low-enriched uranium in the closed dry conversion system requires only routine radiological safety precautions. Spill prevention interlocks / alarms previously described for nuclear criticality safety also enhance radiological safety.

All UF6 vapor-bearing lines to the disentrainment chamber jet are enclosed within pipechases, maintained at a negative pressure with respect to the operating area, and vented through the HF vent scrubber system.

At each end of the kiln, nitrogen-purged lip seats are used to allow rotation of the barrel without loss of process gases or solids. A low-pressure monitor is fitted on the primary section g

'()

of each seal unit and seal purge flows are continuously monitored.' Given a fault condition, these monitors will initiate a safety-related process alarm for corrective action.

Make-break connections at the check hopper / powder transfer container (modcon or geocon) interface are enclosed within Zone 11 containment.

m-

. Occupational Safety and Health Excess pressure-interlocks / alarms previously described for nuclear criticality safety, and excess temperature trips / alarms to prevent damage to costly equipment, also enhance industrial safety.

C)

D-3.39

Hydrogen detection monitors are located above the kiln barrel seals to sample for potential leaks of kiln gases; these monitors will initiate a safety related process alarm for corrective action. Each (previously described) interlock / alarm which shuts off uranium and steam input from the system also automatically shuts off hydrogen input, and introduces nitrogen purge, to the system.

O 15-2.1.4 Recovery of HF for Resale - The conversion kiln off-gas is cooled to recover by-product hydrofluoric acid by condensation.

Off-gas from each conversion kiln flows to its own individual cylindrical condenser (operating 0

at a temperature of approximately 10 C) in which the gases are cooled and HF is recovered as (approximately) 50 w/o aqueous solution condensate.

The condensate is drained continuously from the condenser (alternately) into a pair of cylin-drical HF quarantine ("Q") receiver-sampling tanks in which the condensate is held for anal-ysis of uranium compounds that could result from incomplete reaction in the kiln or from carryover of solids through a failed process filter.

Acceptable low uranium content HF solution is transferred (via an HF pressure tank) to a storage / load-out tank; unacceptable high uranium content HF solution is transferred to a precipitator system in the IDR gaseous effluent treatment area for recovery of the uranium.

O Nuclear Criticality Safety Design and layout of equipment in the HF recovery area enhances nuclear criticality safety of the kiln disentrainment chambers. Controlled vessel overflows and level alarms prevent the HF system from becoming solid with aqueous solution; transfer lines have siphon-breaks to prevent aqueous solution backflow through the system.

Nuclear ciiticality safety within the HF recovery area is (in turn) enhanced by radiological safety design features which provide assurance that only very small quantities of uranium can enter the sensitive items of recovery equipment and which provide backup uranium-in-liquid monitors to give an carly warning of any such uranium entry into the recovery system.

Additional nucle criticality safety assurances arise from the design, dimensions, and operation of the sensitive recovery equipment items. The cylindrical (less than 9.64nch-diameter) con-denser is not designed to hold HF; the condensate drains to the.HF "Q" tanks as it forms.

The cylindrical (less than 9.6-inch-diameter) HF "O" tanks are each designed with a 12-hour

' normal holdup time to permit uranium analysis and transfer to collected acid, either to stor-age loadout or to uranium recovery and neutralization; each tank is equipped with (vent) lines which are arranged to act as overflows, first to the alternate "Q" tank and subsequently to the HF vent scrubber sump (slab with thickness less than 4.4 inches). From a filled HF "Q" tank, O

D-3.40

k recovered acid gravity flows to a cylindrical (less than 9.6-inch-diameter) HF pressure tank and must subsequently be pumped (batchwise) by gas pressure to an elevated head tank (be-fore it can be gravity-discharged through an acid transfer flow leader to the storage / load-out tank outside the building).

e Radiological Safety

. (*]

V.

Under normal conditions of operation, significant quantities of uranium are excluded from the HF recovery system. The uranium hexafluoride and steam supplies on each individual kiln system are interlocked / alarmed by.a high' integrity control system (to assure that in the event

-A of steam failure, the gaseous UF f

6 eed to the system is immediately shut off). Each kiln sys-h tem is fitted with high integrity process filters to remove entrained uranium compounds from the system off-gases; filter integrity is continuously monitored by a differential pressure gauge which is interlocked / alarmed to automatically shut off UF6 (and steam) feed to the system in the event of process filter failure.

As a further safeguard, a continuous uranium-in-liquid monitor is included (in each condenser

. system) which is also interlocked / alarmed to shut off UF6 (and steam) feed to the kiln sys-tem if uranium enters the HF recovery. system.

Recovered HF. is quarantined for uranium analysis before release to the HF storage / load-out tank.

O-a Occupational Safety and Health To protect against the corrosive properties of HF, the HF gaseous and liquid transfer lines are made of Monel or are PTFE-lined and HF liquid holding vessels are PTFE-or rubber-lined.

Operations personnel'are provided with rubber gloves and full face shields as additional pro-tective equipment for manual HF sampling and transfer activities.

Condenser off-gas (containing hydrogen; nitrogen, and traces of HF and water vapor) is trans-ferred to the condenser off-gas scrubber system. From within the HF recovery area, HF vessel

' vent gases (containing traces of HF) are transferred to the HF vent scrubber system; the acid

, p);

(-

transfer flow header provides a means of venting the yard tanks to the HF vent scrubber sys-tem.

~

15-2.1.5 ' UO2 Powder Blending and Storage - The process utilizes bulk powder processing tech-niques for homogenization of UO2 and U 03 8 p wders with CONPOR [(NH )2 2 4 H2 I+Sl C0' 4

i

()f

- additive to produce powder blends for pellet fabrication; the process can also utilize bulk powder processing techniques for enrichment blending of. UO2 and U 03 8 p wders.

The input materials for powder homogenization are UO ; U 0, and porosity control 2

38

-.(CONPOR) additive. These powders are retained in an in-process storage area to await charging n

D-3.41 i,

into an orbital blender. This in-process material storage is provided to accommodate the batch weration of the blenders. Filled, weighed, and identified bulk UO2 p wder containers (trans-ported via lift truck from the dry conversion load out area, the high fluoride powder rework load-out area, and/or received from outside sources) are lifted,via overhead cranc,and stored in the in-process storage area specified for UO ; filled, weighed, and identified bulk U 038 2

powder containers (transported, via lift truck, from the chemically pure scrap recovery load out area or the chemically impure scrap recovery load-out area) are lifted via overhead crane and stored in the in-process storage area specified for U 0 : p reformer blending additive is stored 38 in the in-process storage area specified for CONPOR. The maximum quantities of UO, U 0 >

2 38 and CONPOR additives to be retained in an in-process storage area are 8000 kg, 2000 kg, and 100 kg, respectively.

A pellet pilot test is performed on tSe UO p wder to determine the quantities of recycle U 038 2

to prepare a suitable feed material and CONPOR additive to be homogenized with the UO2 for subsequent pellet fabrication. This test, conducted in the Pilot Pelleting Area, uses represent-ative samples withdrawn from the UO2 nput lot.

i UO2 powder is charged from filled bulk containers for homogenization when one of the blend-ers has become available for powder collection. The filled powder container is mounted on a weigh scale (load cell) via overhead crane at the loading station. The container port is con-nected to the blender system using a flexible connector seal. The powder is then rate fed by a vibratory feeder through a screen and delumper, and is carried by gravity into the blender.

When a predetermined amount (minimum working capacity) of UO2 p wder has been collected in a blender, operation is initiated. The homogenization operation is conducted in an atmos-phere of dry nitrogen. An "on-off" cycle of the blender is utilized; each five minutes of rota-tion is followed by a ten-minute rest period.

Upon receipt of the pellet pilot test results, required quantities of U 03 8 p wder and CONPOR auditive are prepared for homogenization. U 03 8 p wder is weighed and fed into the blender by the same procedure previously descr; bed for UO2 p wder. At a CONPOR additive prepara-tion station, the CONPOR agent has been sampled, analyzed, milled, and sieved and silica base flow promoting agent has been mixed in as preparation for blender loading. This poreformer additive is weighed and transferred into a CONPOR. feed hopper above the blender and the additive is then charged intermittently to the blender during the subsequent homogenization operation.

Upon completion of powder homogenization, representative samples of the resultant mix are taken to ancillary service areas and are quality-control evaluated for physical properties, chem-ical assay, and verification of isotopic enrichment. (Acceptable product wih be approved by quality control for fabrication operations; off-specification product will be consigned to rework processes.) Based on results of the blend analytical parameters, the homogenized powder can be D-3.42

4 rate-fed directly from the blender system to the pelleting area or can be discharged from the

\\

blender system directly to bulk containers for on site storage and subsequent praessing, or into appropriate containers for off-site shipment.

The blenaer system can also be used to mix and transfer batches of uranium powders (UO2 and U 0 ) of different enricoments (up to 5 w/o U-235), using applicable procedures pre-38 viously described for powder homogenization. Upon completion of enrichment blending, the mix is quality-control sampled and analyzed. The blended powder can then be prepared for release to the pelleting area or can be discharged to bulk containers for on-site storage and subsequent processing, or placed into appropriate containers for off-site shipment.

Nuclear Criticality Safety m

1.

Moderation Control Contingencies The IDR blending operations rely principally on moderation control for nuclear criticality safety. Operations are designed such that multiple contingencies (accidents, failures, or events) must occur before conditions for a criticality accident are possible. The systems are evaluated by postulating contingencies which could result in sufficient moderation getting into suffi@. enoched uranium in an unsafe geometrical configuration and by addressing the rois which prevent these events from occurring. (At least double-D.

contingency criteria are satisfied for each scenario.)

- V Scenario: Sufficient Moderator Enters Blender as Moderated UO2 and/or U 038 This is simply a special case of the " Sufficient Moderator in Moderation Controlled (Moc' con) Container'? scenario previously discussed in the Dry Conversion nuclear critical-ity safety analysis. It is postulated that the same failures occur. (However, it is further postulated that insufficient moderator gets into a single container, but is accumulated in

-two or more containers, such that when these containers are all loaded into the blender

~

2 s sufficient to cause a criticality condition.) Thus, i

the moderator and quantity of UO the same preventive contingencies apply; however, there is an additional contingency in that all failures required for. sufficient moderation in a single modcon container must

' v occur, for two or more containers.

i -

Scenario: Sufficient Moderator in ~ Blender Prior to Filling with Dry ljO2

.The blenders are located within carefully ' segregated (" water barrier") moderation control

)

areas where moderators are strictly limited by administrative procedures; thus, these

~'

procedures must fail before moderator could get into an empty blender. Further, inspec-

. tion procedures require checking the blender before sealing F.d/or filling; thus, these

~

procedures also must fail before UO2 could be put into a blender holding moderator.

A

.~Iu)

D-3.43

Scenario: Sufficient Moderator Enters Blender Filled With Dry UO2 Administrative procedures strictly limiting moderating materials in moderation control areas, or the control area water barrier itself, must fail before moderating materials could enter the moderation control area. Then, administrative procedures governing the opening of the blender must fail, or the blender integrity itself must fail, before any such avail-able moderator could enter the blender. Further, these administrative controls and/or barriers would have to fail within a short time of each other in order for the potential criticality condition to occur (otherwise, the administrative procedures requiring prompt corrective actions would also have to fail).

O Scenario: Sufficient Excess Moderating Additive Put in Blender The CONPOR addition system is carefully designed and operated to limit the maximum quantity of CONPOR to a single batch, to prevent over batching, and to rate-feed the additive during blending to minimize local concentratic,.s. To get excess CONPOR into a blender of UO, first the administrative procedures controlling the receipt and addition 2

of CONPOR must fail, then an additive interlock / alarm system must also fail, thus per-mitting the administrative failure to result in charging excessive CONPOR into the blender.

The blender was modeled in KENO as an optimized configuration which minimizes neutron 1:akage from the system while maximizing the effect of reflectors (both water and con-crete) external to the system. Additional model input included 5 w/o U-235 enriched uranium with the maximum moisture content of 0.3 w/o H O and a double batch of 2

CONPOR additive. Further, the uranium and moderator were modeled to mix in a sphere sized to result in optimum moderation of the UO, with the sphere located in a position 2

to maximize the reflection by surrounding dry UO. The calculated keff was still less 2

than 0.95.

Scenario: Sufficient Wrong (Moderating) Additive Put in Blender Administrative procedures controlling procurement of moderating additives must fail in order for the wrcng additive to be supplied. Then, the administrative procedures for additive sample analysis must subsequently fail to reject such wrong additive before it could be added to the UO2 n the blender.

i Scenario: Blender Internal Gearbox Failure With Release of Sufficient Wrong Lubricant Special lubricant with low moderating properties is used in the blender internal gearbox.

To have the wrong lubricant (i.e., one with high moderating properties) in the gear box, a wrong material must be supplied and the administrative procedures for !ubricant sample analysis must fail to reject this wrong material. Then, the filled gearbox must subsequently O

D-3.44

i tl' fail before a sufficient quantity of the wrong lubricant could be released to the UO2" I

the blender.

4 2.

Safety Interlocks Although moderator (CONPOR additive) is introduced into the blender for homogeniza-

)

- tion with dry uranium powders, the quantity of moderator added and the procedure for addition are carefully controlled to ensure that the system remains undermoderated. Nu-l clear criticality safety control has been directed to the prevention of significant modera-F

. tion, since technical process parameters require blender sizes in excess of those required for geometry control.

,w

~ Assurance that excess moderating additive cannot be introduced into the blender is based f-4~

on a detailed interlock / alarm system incorporated in the design to automatically prevent such introduction. Another detailed interlock / alarm system is incorporated in the design

{

to detect the presence of tr.oisture in the blender system and to automatically apply corrective actions well before potentially hazardous conditions could develop.

i t

t Further, rigorous attention is applied to excluding moderators from the secondary water 1

f' barrier (the room) and from moderation-controlled (modcon) bulk containers. As con-

[

tingents to this effort, additional interlocks / alarms are installed to prevent inadvertent t-

. discharge and/or spills of. moderation-controlled powder from the primary water barriers E

(blender and modcon containers) and to prevent inadvertent entry of prohibited mod-y-

erators to the primary water barriers.

~e

' Blereder; Feed System interlocks.

g

- The following interlock / alarm in the CONPOR feed s/ stem will automatically prohibit CONPOR l

t p

7 addition to the blender upon detection of the indicated fault:

a A-

[

' 1, Attempt to add more than a single batch of CONPOR to the feed system before the feed -

h

' system has emptied from tM last previous CONPOR addition and subsequent total blender

. discharge is complete.

l' /G

!d Thel following interlocks / alarms in the UO /U 0 3 8 eed system will automatically prevent dis-f 2

I'

~

- chargeL(an'd/or spills)'of powder;to the blender (or the room areal upon detection of the in-

' dicated fault:

fr~

1.

Inadequate uranium feed container connection -

i; A j L 2.

Inadequate blender outlet closure f 3.'

> Overfill of blender D-3.45 r -.

[.

w,

,... =,-

~-,..=.-...,..a..-.-.==-

= -.. -.. -.

o Blender Rotation Interlocks The following interlock / alarm in the blender rotation system will automatically prohibit blender operation upon detection of the indicated fault:

1.

Inadequate blender (inlet and outlet) closure The following interlock / alarm in the blender rotation / nitrogen feed systems will automatically prohibit blender operation and shut off nitrogen flow upon detection of the indicated fault:

1.

Moisture in blender purge gas.

O Blender Discharge System interlocks The following interlock / alarm in the homogenized uranium powder discharge sys' tem will auto-matically prevent discharge (and/or spills) of powder from the blender (to the room area) upon detection of the indicated fault:

1.

Inadequate blender inlet closure 2.

Inadequate blender - press feed hopper interface (or discharge container) connection 3.

Overfill of press feed hopper (or discharge container).

3.

Ancillary Controls All other process vessels and equipment (i.e., except for the blender and the UO /U 038 2

modcon containers previously evaluated) in the moderation-controlled powder storage and blending area normally containing uranium or having the potential to contain ura-nium as a result of accident conditions are designed safe by geometry for 5 w/o U-235 materials under full reflection and optimum moderation. The CONPOR feed hopper is further designed to hold a maximum of a single batch to provide an additional contin-gency against excess batching conditions.

bulk (modcon) containers are stored in a one-high planar array with a Filled UO /U 038 2

minimum of one foot edge-to-edge container separation, thus they require no interaction considerations; in the powder storage ageas, fixed steel racks maintain the required spac-ing. It is not considered possible that fault conditions could arise to such an extent as to cause interaction between materials, either with respect to adjacent units wahin a single blender system or between adjacent blender systems.

O O

D-3.46 f

a Radiological Safety Low-enriched uranium in the closed powder blending system requires only routine radiological safety precautions. All uranium powder storage containers are fitted with efficient sealing lids which are kept tightly closed unless the container is in a fill or discharge position. Spill O

prevention interlocks / alarms, previously described for nuclear criticality safety and excess temperature and pressure interlocks / alarms to be described for occupational safety and health, also enhance radiological safety.

- The following processing sections of the blender system are enclosed within Zone 11 contain-ment:

1.

UO2 and U 03 8 charging ports to the blender system 2.

- Uranium vibratory feeder and delumper 3.

Uranium discharge from the blender system to the pelleting area (or powder containers).

The process off-gas (blender nitrogen purge) passes through a roughing filter followed by a HEPA filter prior to discharge into the plant ventilation system.

Occupational Safety and Health m

Transfer and handling of UO /U 038 bulk containers is in accord with applicable lifting device 2

O'

_ safety practices embodied in Site Regulatory Compliance Manual Procedure RCM-501.

~

- The internal blending action is conducted in an atmosphere of dry nitrogen which prevents oxidation of the uranium powder at temperatures induced by the mechanical energy input;

the "on-off" cycling of the blender prevents excessive powder temperatures. The following interlocks / alarms in the. blender rotation system will automatically prohibit blender operation

- upon detection of the indicated fault..

1.

Low blender purge gas flow 2.

High blender temperature 3.

High blender pressure.

CONPOR agent and additive is stored in sealed containers when outside confinement or con-tainment. The following sections of the blender system are enclosed within Zone il contain-ment.

1.

CONPOR additive' preparation station V'

22.

CONPOR ' rate feeder to the blender.

T'N A_)

D-3.47 -

15-2.2 Fabrication Following homogenization in the Conversion Area, the blended uranium oxide powder is (closed system) transferred to the Fabrication Area where it is compacted, granulated, and pressed into pellets. Pressed pellets are loaded into boats, which are subsequently charged to electrically heated sintering furnaces, where they are transformed into high density fuel pellets by sintering in a reducing atmosphere. Sintered fuel pellets are then processed through a grinding operation to obtain specified dimensions. Ground pellets are loaded into prepared metal tubes, plugs are inserted, and the resultant fuel rods are hermetically sealed by welding.

Finished rods are inspected, tested, then transferred for final assembly. Here, the fuel rods are loaded into designated positions in a prefabricated support structure (" skeleton") consisting of thimble tubes and structural grids. Top and bottom nozzles are then attached to com-plete the final fuel assembly.

These fabrication operations, which are described in detail in the following subparagraphs, represent (in essentials of equipment and procedures) proven processes which have been carried out safely and economically for some 11 years at the NFD Manufacturing Depart-ment's Columbia Plant (under License SNM-1107, Docket 70-1151). The success of the Columbia operation, while not a subsititute for the detailed analyses to follow, is strong sup-porting evidence of the adequacy of the safety controls which are designed into these operations.

15-2.2.1 Fuel Pellet Fabrication - Homogenized (and quality control released) uranium oxide powder is fed directly from the blender discharge to the powder compactor feed hopper on demand as controlled by an automatic level sensor in the feed hopper. The powder is then compressed by a (two-stage) compactor press at parameters determined by prior pellet pilot testing to produce a compact suitable for granulation. Pellet pilot testing is conducted in a pilot pelleting area where test pellets are fabricated from samples of input powder to the p:llet fabrication lines, using process steps and controls similar to those described here for those production lines. Compacted powder is next comminuted in a granulator; then, result-ant granules, held and fed from a feed hopper, are proportionally mixed with (binder /

18352)2] in a roller to produce a consistent quality lubricant) zinc stearate (Zn (C H 0 press-feed material. Pellet press-feed material is discharged to a press hopper for subsequent compaction into " green" (unsintered) pellets by a pellet press. Resultant pellets are mechan-ically removed from the press table in an orderly array and are loaded into sintering boats.

The press operator performs periodic measurements on these pellets. Green (unsintered) scrap, along with chemically pure uranium oxide powoer, can be recycled back into the process stream prior to granulation.

O D 3.48

i-l

[

The filled-sintering boats are moved along a conveying system to the electrically heated smtering furnaces. Storage for the boats is provided by the length of this conveying system.

. The pellets' are sintered in a (hydrogen) reducing atmosphere. From prior pellet pilot testing rasults, prccess information is generated for each powder blend which specifies the time cycle s

~ and sintering temperature. As each boat exists a sintering furnace, representative pellet samples

[

~

are taken; the sample pellets are inspected by Quality Control and, if tests are acceptable,

~

' the represented pellets are released for dry grinding. Rejected or excess sintered materials are cent to the chemically pure (" clean") scrap recovery area for oxidation to powder and recycle back into the process stream.

{

N i

Quality Control released boats of sintered pellets are unloaded into a pellet orienter for feed-(-

ing.into the grinder, where the. pellets are ground to specification. Pellets are manually sampled and checked by the grinder operator to assure that the process is in control. Finished pellets are then loaded into trays,lwhich are placed in carts and moved to storage to await final

~

-inspection. Quality Control samples and analyzes ground pellets and if they are acceptable they are released for loading into fuel rods. Most materials collected from the grinding of

. sintered pellets are sent to the clean scrap recovery area for oxidation to powder and recycle

~

back into the process stream;. pellets soiled with oil,' grease, etc.; along with uranium oxide u

powders containing tramp materials, are sent to the chemically impure (" dirty) scrap recovery j f --

)

' area for reclamation to pure uranium oxide powder prior to recycle back into the process

((

stream.

L

a Nuclear Criticality Safety Although all special ~ nuclear material in the-fabrication processes, from blend release to pellet

~

i tray; storage,.is verified ' dry (H/U <0.3); each step described is controlled using subcritical g

limits'which assure nuclear criticalit_y safety under postulated conditions of. moderation and

" reflection.

1Beginning-withf the (<9.6-inch-diameter) compactor hopper, through the (<9.6-inch-diameter) roller. feed and pellet press hoppers,' to the pellet' press, suber'tical limits for homogeneous oxides are utilized. Beginning with' 'the pellet press, including the (<3.9-inch slab /<8.7-inch.

_ equivalent diameter) sintering boats, through the. pellet tray storage units, subcritical limits 1-

' for heterogeneous oxides are utilized.-

' Granulated uranium oxide powder is mixed with binder / lubricant in proportions insufficient-to' increase the H/U beyond'0 3; mixing is accomplished by rolling a (<9.6-inch-diameter and/or <26-liter volume) container [ holding the' powdered: ingredients.

l The movement and storage of loaded sintering boats in the furnace area, including associated conveyors,- are controlled. on_ a slab or cylinder basis by the depth or cross-sectional. area f

l D 3'.49.

s

of the boats. Physical and/or administrative controls are utilized to prevent the stacking of loaded boats beyond the (<3.9 inch slab /<8.7 inch-equivalent diameter) subcritical limit.

For grinding operations, '>ellets are fed and unloaded on a <3.9-inch subcritical slab limit basis.

Finished pellets are stored on special pellet trays in storage units. Each pellet tray storage unit provides for <5 slabs separated vertically by a (minimum) 12 inch spacing; each slab being 4 trays (<3.9 inches) thick. Pellet trays and their associated storage units are designed not to permit retention of water.

Off specification pellets (scrap) are accumulated in (<27 kg U) subcritical mass limit batches, or (<3.9-inch) subcritical slab limit containers, pending transfer for recovery.

The pellet press hopper spacing area of >8.0 square feet provides a " smeared" slab thickness less than the (1.3 inch) maximum controlled parameter value. Other spacing requirements are generally satisfied by the fact that any sir:gle hood or unit of processing equipment is normally limited to one safe subcrit. More than one subcrit is permitted only when each subcrit is pro-vided with, and maintained within, its unique safe nuclearly area. Each hood or equipment unit occupies a floor area of from 12.5 to more than 25 square feet. Adjacent pellet tray storage units are arrayed to form continuous slabs.

o Radiological Safety Low-enriched uranium in the closed fuel pellet fabrication system requires only routine radi-ological safety precautions. All uranium storage containers are fitted with lids which are kept in place unless the container is inside of, or securely connected to, Zone 11 containment.

The following processing sections of the fuel pellet fabrication system are enclosed within Zone il containment:

1.

Compactor press, granulator, roller, pellet press 2.

Boat unloader 3.

Pellet orienter, pellet grinder.

Ventilation systems servicing the following processing sections of the fuel pellet fabrication system pass through separators for recovery of entrained uranium values before HEPA filtra-tion and discharge into the controlled area and/or the plant ventilation system:

1.

Compactor press, roller, pellet press 2.

Boat unloader, pellet orienter, pellet grinder, pellet cleaning unit, pellet tray loader.

O D-3.50

Occupational Safety and Health a

Excess temperature trips / alarms installed in sintering furnaces to prevent damage to costly equipment also enhance industrial safety.

1A Sintering furnaces are purged with nitro 9e prior to introduction of hydrogen. Hydrogen b

detection monitors are placed at representative locations above the sintering furnaces to sample for potential leaks of furnace gases; these monitors will initiate a safety-related process alarm for corrective action. A hydrogen pressure interlock / alarm is provided for each furnace, which, upon indication of hydrogen pressure below the minimum setpoint, automatically shuts off hydrogen input and introduces nitrogen purge to the system. To present leakage of hydrogen to room atmosphere, sintering furnace entrance and exit doors are protected by natural gas flame curtains (with automatic pilot lights).

15-2.2.2 Fuel Rod Fabrication and Storage - Pellet trays are removed from their storage units to prepared rod loading fixtures. The fuel pellets are transferred from the trays and loaded into empty fuel tubes. Loaded tubes are cleaned, plugged, placed on a conveyor, and transferred to weld stations where they are inserted into a fixture and (laser) welded.

Finished fuel rods are then subjected to a variety of inspection operations including radiog-raphy, to assure the quality of the final product. Quality Control released rods, segregated into fuel assembly lots, are stored in channels, with one fuel assembly complement per m

' channel, awaiting final fuel assembly; rejected rods from all inspection operations are trans ferred to the storage area where they are inaded into channels to await corrective action.

. Rejects are subsequently moved to the rod repair area where the ds are simply repaired or where they are opened (using drill and lathe) and expanded (using high-pressure water) so that the contained fuel pellets can be salvaged for recycle to the process.

.a Nuclear Criticality. Safety Rod loading operations are all controlled on a (<3.9-inch) slab basis in which usually one pellet or rod diameter is a practical slab thickness; quality assurance operations are controlled

.p on a (<3.9-inch) slab basis; rod repair operations are controlled on a (<3.9-inch) slab or h

(<8.7-inch diameter) cylinder basis.

I Close-packed fuel rods in channels are stored throughout the plant as slabs with a thickness l

less than, or equal to, the MPV of 15 inches.

O e

Radiological Safety

.V j

Low-enriched uranium in the closed rod loading and repair systems requires only routine f

radiological safety precautions; quality assurance operations are performed in uncontrolled areas.

-O l X)

D-3.51 L

The drill / lathe in the rod repair area is provided with Zone 11 containment.

Appropriate shielding and procedures are provided for radiography operations.

o Occupational Safety and Health For the rod welding operation, appropriate operator protection and procedures are provided to prevent eye injury from lasers.

In the repair area, rods are drilled at the top end plug to safely relieve internal pressure prior to subsequent operations. Also in this area, high-pressure water (used to expand the fuel tubes to enable pellet salvage) represents significant stored energy which is suitably contained to pre-ciude a sudden release of the energy from endangering operators or equipment.

15-2.2.3 Final Fuel Assembly Fabrication and Storage - Finished fuel rods are removed from their channels, cleaned, and loaded into rod magazines. The magarines are transported via cart to a loader and fixture, where the fuel rods and fixture skeletons are mated into final fuel assemblies which are subsequently transported via crane. The fuel assemblies are inspected and cleaned (using appropriate materials); then, the finished fuel assemblies are stored pending shipment.

C Nuclear Criticality Safety Except for the cleaning steps, fuel assembly fabrication operations are performed dry; com-pleted fuel assemblies are handled in a manner which provides a minimum 12 inch, eJge-to-edge spacing.

Final fuel assembly cleaning operations are performed in three submerged wash tanks, each with assembly spacing features. The center-to-center spacing between loaded assemblies is maintained >30 inches, and a minimum of 12 inches of water is maintained between moder-ated assemblies. The wash solutions provide simultaneous isolation and moderation; thus, moderation without isolation is not credible for the open assemblies. The MPV applied is assembly reactivity (keff + 20 <0.95). The wash tanks contain no strong corrosive agents capable of dissolving the fuel rod cladding and permitting a more reactive arrangement of the contained nuclear fuel.

During transport, at least a 12-inch, edge-to-edge separation is maintained between fuel assemblies.

In final fuel assembly storage a miriimum of 12-inch, edge-to-edge assembly separation is maintained.

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8 I_IO a

Radiological Safety

.O Low-enriched uranium contained in final fuel assemblies requires minimal radiation protection precautions; assembly fabr. cation, inspection and storage are conducted in uncontrolled areas.

_a Occupational Safety and Health i.

Transfer and handling of loaded fuel assemblies via overhead crane is in accord with applicable t

h lifting device safety practices embodied in Site Regulatory Compliance Manual Procedure RCM 501.

I 15-2.3. ' Ancillary Services i.

The conversion and fabrication operations just described, and the scrap recovery and waste treatment operations yet to be described, require a variety of ancillary support services.

intermediate products (e.g.,-uranium powder or pellets) might be received for introduction at

^

appropriate points in the process for follow-on fabrication operations. Samples are received by

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the analytical laboratory for chemical, radiochemical, physical, spectrographic, and metallo-graphic analyses. Protective clothing, respirators, towels, etc. (from SNM Building operations) are laundered for. reuse. Completed. fuel assemblies and other plant fuel products are loaded into approved (NRC/ DOT) shipping containers for delivery to utility reactor sites, etc. Such

' ancillary services are evaluated in 6 tail in the following subparagraphs.

'15 2.3.1, - Receiving intermediate Fuel Products (From Outside Sources) - Uranium oxide is

[:

received from outside suppliers in licensed packages. These packages are opened inside the

- SNM Building, and the contents are transferred to bulk containers (sized and loaded to plant

.. specifications,~ which are weighed,' identified, transported, and stored (as described in sub-

. aragraph 15-2.1.5). Receiving uranium. oxide is typical of licensed material receipts.

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-a Nuclear Criticality Safety.

' Individual shipping / receiving' packages are sized to assure nuclear criticality safety under normal

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i D-a handling and credible-accident conditions.

)'j Samples of received special nuclear material (SNM) are taken and sent to the analytical

' laboratory for testing to' verify the U-235 enrichment, dryness-(H/U < 0.3,' as prescribed by-l' imaterial. orders)f and other specifications prior to interim-storage and/or transfer of the

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I contents to bulk UO2 powder containers; Uranium oxide powders' (enriched 45.0.w/o in U-235) are ' received in <9.6-inch-diameter b

' inner containers which may'be stacked in coaxial cylinders while awaiting contents transfer.

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Interaction between'such cylinders. is not of concern due to the dryness-(H/U _<0.3) of the powder; however/ the containers are placed on fixed steel storage shelves ' esigned to assure d

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that the containers on the shelves form vertical columns of long cylinders with their axes parallel and that these columns are spaced a minimum of 12 inches edge-to-edge.

o Radiological Safety The powder containers are closed and sealed when received. They remain closed until the sampling and transfer steps and therefore require only routine radiation protection precautions.

When they are cpened for sampling or transfer of contents, they are within (or securely fastened to) Zone il containment.

G Occupational Safety and Health The fixed steel shelves for interim container storage are designed to retain their integrity against failure under credible loads, shocks, or collisions.

15-2.3.2 Analytical Laboratory - Samples of special nuclear material (SNM) which have been selected at various stages during fuel conversion and fuel assembly are analyzed in the labora-tory. Established techniques for wet chemical, spectrographic, and other types of analysis are utilized. Samples and/or samples of materials from other Westinghouse operations, including samples of received intermediate products for fuel fabrication processing and/or recovery of uranium values, and from other suppliers (including samples for verification of correct non-SNM additives to the process) are also received and analyzed.

Uranium-233 and uranium 235 are used in the preparation of precisely controlled, standard solutions. Samples of these solutiens will be used primarily for instrument calibrations and frr specific analytical tests (for example, U-233 solution is used for the determination of total uranium in Zircaloy by the isotopic dilution method).

O Nuclear Criticality Safety The total quantity of SNM required in the laboratory is small and is controlled so that no laboratory room holds more than 700 grams of contained U-235, which is the maximum permissible mass at any enrichment that is not in itself subject to the criticality monitoring requirements of 10CFR70.24.

The mass limit for an individual laboratory room might include up to 5 grams of U-233 (the maximum license possession limit), so for'g as the total contained U-235 plus U-233 in that room does not exceed 700 grams.

o Radiological Safety Low-enriched uranium analytical activities require only minimal radiological safety precautions.

Most of the samples are handled in solution form. When preparatory operations with dry samples must be performed (such as weighing out uranium powder samples or crushing O

D-3.54

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sample fuel pellets prior to dissolution), they are conducted in HEPA-filtered Zone ll contain-ment; operations with U 233 are limited to individual (maximum) 1-gram quantities in a HEPA-filtered Zone li containment.

m' Occupational Safety and Health Q("^s Analytical activities in support of low enriched uranium fuel fabncation operations require only routine industrial hygiene practices normally associated with the prudent operation of industrial laboratories. These practices include use of lab coats and safety glasses, and provi-sion of fume hoods to contain volatile chemicals, dissolutions, powder aerosols, etc., emer-O gency showers and eye washers.

Cl 15-2.3.3 Laundry - At initial facility operating capacity, it is currently planned to send pro-tective clothing, respirators, towels, etc. from SNM Building operations to an approved service company for laundering prior to reuse. Future installation of an in-plant laundry will most likely incorporate a dry cleaning process in which nuclear criticality safety is based on mass control and radiological safety (and occupational safety and health) is based upon routine radiation protection measures and routine industrial hygiene practices conducted by trained, technically competent, personnel.

15-2.3.4 Shipping Final Fuel Assemblies - Final fuel assemblies are transported via crane and loaded into NRC/ DOT-approved, and Quality Control inspected shipping containers, two

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fuel assemblies per container. Loaded shipping containers are securely closed using approved procedures. Quality control verifies the integrity of the loaded shipping containers; then, the containers are transported via crane for placement on the shipment vehicle or to interim storage awaiting shipment. (Fuel assemblies may be stored, loaded in shipping containers outside the SNM Building but within the Plant Area security fence.) Shipping of final fuel assemblies is typical of licensed material shipments.

Nuclear Criticality Safety-a During transport for loading, the cranes assure at least a 12-inch, edge-to-edge separation between fuel assemblies.

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Individual shipping / receiving cor. tainers are designed to assure criticality safety under normal handling and credible accident conditions.

' Radiological Safety m

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Low-enriched uranium in final fuel assemblies requires minimal radiation safety precautions; assembly shipping coritainers are loaded and stored in uncontrolled areas.

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o Occupational Safety and Health Transfer and handling of final fuel assemblies and loaded assembly shipping containers is in accord with applicable lifting device safety practices ernbodied in Site Regulatory Compliance j

Manual Procedure RCM-501.

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15-2.4 Scrap Recovery (To be supplied)

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15-2.5 Waste Treatment (To be supplied)

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I 15-3 ORGANIZATION AND ADMINISTRATION Safety means an acceptable balance of risk against benefit; it is meaningless as a concept isolated from other goals. It thus follows that safety must be considered an integral goal of design and operation instead of something superposed. Safe operation and accident pre-b) vention depend upon delegation of responsibility and authority for safety implementation to V

the properly qualified supervisory level closest to the actual operation, under the general direction and policies set by management, and with advice and guidance from qualified safety experts. This is the basis of safety related organization and administration at the Alabama Nuclear Fuel Fabrication Plant (ANFFP).

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15-3.1 Organization Chart The organization chart shown in figure 15-16 indicates the administrative relationships of the components and disciplines of interest in this license application. 3.2 Functions of Organizational Units The Westinghouse Nuclear FO. Division (NFD) Manufacturing Department will operate the fuel fabrication facility at Prattville, AL for Westinghouse Nuclear Energy Systems (NES).

Within the NES Water Reactor Divisions (WRD), a Licensing, Safeguards & Safety group con-ducts a variety of activities which primarily provides services to other divisions. NES Nuclear

'V Materials Management & Safeguards and NES' License Administration are both members of this group.

The Nuclear Fuel Division, which is a division within WRD, is responsible for the manufacture of fuel-bearing components for commercial nuclear reactors. NFD includes a highly qualified engineering group which is responsible for assigned fuel design and development projects, qualified components within this group provide nuclear criticality safety calculation services.

However, the Nuclear Fuel Division's primary function, to fabricate ar:d ship commercial reactor fuel assemblies, is the. responsibility of the NFD manufacturing group. NFD manufac-turing (through a Plant Manager reporting to the Division Manager) will operate the Alabara V

Nuclear Fuel Fabrication Plant (ANFFP) as'a significant portion of its activities toward carry-ing out the assigned responsibility to fabricate Special Nuclear Material (SNM) from various raw material forms into finished fuel' assemblies. In addition, NFD manufacturing directs

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support activities at the ANFFP site and is responsible to provide all required Operations f( 3 assistance functions such as Planning, Facilities, Production Control, Administration, Con-U troller, Personnel Relations, and Regulatory Compliance.

Consultation, general surveillance, and coordination of radiological safety and inaustrial hygiene practices among the various Westinghouse sites is provided by an industrial Hygiene

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' Group which functions at the Headquarters staff level.

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15-3.3 Functions of Management Positions Within Organizational Units Each Westinghouse management position is covered by a written description which spells out

'in detail-its scope, purpose, duties, responsibilities, authorities, and requirements. The descrip-tion clearly separates the incumbent's authority for decisions which he may make unilaterally from those requiring higher management approval. It defineates relationships with other func-tions. It specifies responsibility for personnel and for control and maintenance of facilities i

and equipment. Each such position description is reviewed and approved by two higher levels

' of. management.

The NES Management Position Committee, which is made up of key members of the NES I

staff, reviews and evaluates these positions. Such reviews determine that all key functions are covered, that interrelationships are clear, and that conflicts are eliminated.

Individuals are selected to fill these management positions by evaluating their capabilities to perform the various activities specified in the position description. Two higher levels of l-management (as a minimum) must approve each selection (or change) of an incumbent. Con-l tinuing high level performance of these individuals is assured through a formal program of annual review.

p Operation of'the Manufacturing Department of the Nuclear Fuel Division is in accordance tI with ~ the general. operating philosophy and procedure employed in all Westinghouse plants and k(

facilities.' Briefly,-this philosophy provides that responsibility for all phases of operations, in-cluding safety and health protection, follows-the usual lines of organizational authority. (That is, supervision is made as responsible for safety and health protection as for production. This points out1th'at accountability' for. safety and hcalth differs ia no intrinsic way from account-

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ability for operations and that good managerial practices apply to both.)' Advisory technical experts and service groups arelprovided to assist'line management in the evaluation of opera-g

tions within their control and to provide measurements, determinations, and ir$ formation which '

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. aid'in the analysis of specific' operations and situations. Such advisory and service functions b

in no..way. relieve individual line managers from responsibility for the safe operation of their-

'f-functions and facilities 'or for ascertair.ing"and assuring.(through ~ appropriate management:

I.

y channels) that adequate advice' and service is provided Basic policies and procedures are i

Testablished by line. management'with'the approval of cognizarit staff groups; and within the p;

framework 'of. these policies,uthe responsibility for. making decisions at the ' operating. level bi f rests;with the firs't'ievel. managers (shift supervisors).?They have a basic responsibility-to per-

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fo mitheir assigned' operations in a safe and. orderly manner.

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The shift supervisors are responsible for. providing operating instructions for the guidance and

direction of nonmanagement' personnsi under their direction. Written procedures are ~ prepared -

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and reviewed by' technical experts. These procedures become.the bases for performingLspecific ^.

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operations. The shift rJpervisor cannot make unilateral changes in such written procedures independent of cognizant technical experts.

The shift supervisors are also responsible for assuring that the personnel under their jurisdic-tion receive adequate training. Cognizant Regulatory Compliance personnel fint present an orientation lecture to new employes. Bio'ogical effects of radiation, measurement and control of radiation exposure and radioactive material, the radiation protection program, and emer.

gency preparations we the topics discussed. To acquaint the new employe with basic reguia-tions, several parts of Title 10, Code of Federal Regulations, are presented. Emphasis is placed upon 10 CFR, Parts 19 and 20. The cognizant shift supervisor then assigns experienced employes the responsibility of indoctrinating and training the new employes in the proper procedures and precautions in performing their specific jobs. The shift wpervisor evaluates tna progress of each new employe; and gradually increases it'dividual job assignments until th) employe fulfills the requirements of the job description. Failure to achieve minimum per-formance requirements is cause for a change in a'ssignment or release. Periodic reinforcement instruction is conducted on a job by the employe's supervisor and/or cognizant personnel from Regulatory Compliance. Changes in regulations, operating conditions and/or procedures are promptly relayed; administrative policies are reiterated.To ascertain that all employes remain familiar with the building emergency evacuation procedures, drills are conducted. After cach drill is evaluated, appropriate management is informed of any shortcomings and they instruct personnel under their supervision regarding any clarifications required.

15 3.4 Functions of Regulatory Compliance Positions Within Organizational Operating Units A staff activity with responsibility for regulatory compliance maintains a general surveillance program and associated reports system and instigates remedial actions (at appropriate manage-m;nt levels) as required.

Operational Health Physics responsibilities are exercised by a Health Physics Manager who directs the radiological surveillance programs described in the Regulatory Compliance Manual.

Specific assignments are delegated to health physics technicians who have received documented training in the following subject areas: radiation physics, measurement techniques, radiation biology, current regulations, radioactive material handling, shipping, receiving, and waste disposal, laboratory monitoring, and practical exerce 1 and demonstrations.

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D-3.140

The Radiological and Environmental Services group at ANFFP functions on a site-wide basis to assure consistent, coherent, and comprehensive planning and application of overall safety and health programs. Specific assignments are delegated to qualified engineers having the i

education and experience described in Section 4 of the license specificat ons.

In the conduct of Alabama Nuclear Fuel Fabrication Plant activities, any and all personnel involved in the operation of the facility have the continuing right to review or request review of the safety of an operating process or procedure. Further, responsible Regulatory Compliance staff members on duty have the authority to forbid, through the cognizant supervisor, any p

operation which in their expert opinion involves undue immediate hazard, until the situation V

is reviewed with cognizant supervision and there is a satisfactory resolution of remedial methods and procedures to be used. Continuous daily contact, between the personnel of Regulatory Compliance and operational staff members assures appropriate consideration of unique challenges as they arise.

15 3.5 Education and Experience of Key Personnel The qualifications of key personnel assigned to the managerial and technical positions described in this license are:

Director; NES Licensing, Safeguards & Safety (Monroeville, PA) - Provides licensing, e

O safeguards and safety services to ANFFP.

V (Resume to be provided)

Cognizant Component; NFD Engineering (Monroeville, PA) - Provides nuclear a

criticality safety calculation _ services to ANFFP.

(Resume to be provided)

Manager; Headquarters Industrial Hygiene (Churchill, PA) - Provides industrial e

hygiene and general radiation protection services to ANFFP.

(Resume to be provided)

.fm lj Plant Manager. (ANFFP) - Overall responsibility for all activities of the NFD e

Manufacturing Department at the ANFFP site, including those relating to health, safety, safeguards, and environmental protection.

' (]

(Resume to be provided)

'y-Manufacturing Engineering Manager (ANFFP) - Standing member of the Regulatory m-Compliance Review /ALARA Committee.

v D-3.141

(Resume to be provided) e Operations Man.

'^NFFP) - Standing member of the Regulatory Compliance Review /ALARA Co.rimittee.

(Resume to be provided)

C Regulatory Compliance Manager (ANFFP) - Specific administrative responsibility for licensing, health, safety, safeguards, environmental protection, and compliance; standing member of the Regulatory Compliance Review /ALARA Committee.

(Resume to be provided)

Radiological and Environmental Services Manager (ANFFP) - Delegated programmatic o

responsibility for licensing, health, safety, safeguards, environmental protection, and compliance.

(Resume to be provided)

Criticality Engineer (ANFFP) - Delegated implementation responsibility for the o

nuclear criticality safety program.

(Resume to be provided)

Health Physics Engineer (ANFFP) - Delegated implementation responsibility for the a

radiological safety and environmental protection programs.

(Resume to be provided)

Health Physics Manager (ANFFP) - Delegated implementation responsibility for O

operational surveillance aspect of the radiological safety and environmental protection programs.

(Resume to be provided)

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O D-3.142

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