ML19350C415
| ML19350C415 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 01/09/1981 |
| From: | Sabo A WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Page R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML19350C414 | List: |
| References | |
| 18234, WRD-LA-276, NUDOCS 8104010488 | |
| Download: ML19350C415 (36) | |
Text
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WRD-LA-276 O
Westinghouse Water Reactor ea 355 Electric Corporation Divisions PittsburghPenreyNane15230 January 9, 1981 U.
S. Nuclear Regulatory Commission Office of Nuclear Material Safety & Safeguards
' Division of Fuel Cycle & Material Safety Washington, D.C.
20555 Attn:
Mr.
R. G. Page, Chie f Uranium Process Licensing Branch
Subject:
Transmittal of Application for Amendment to Upgrade Facility, License SNM-1107, Docket 70-1151 Gentlemen:
The Westinghouse Electric Corporation hereby requests an amendment to License SNM-1107 to authorize operations with special nuclear material in an upgraded section of our Columbia Facility, in accordance with the attached application.
The information included with this transmittal consists of ten (10) copies of environmental information and altered license conditions (submitted as changed pages, in accordance with applicable license specifications). Je will appreciate your tinely review of this information such that the installation schedule for the proposed line can be maintained.
l Please find enclosed a check payable to the U.
S. Nuclear Regulatory Commission, in the amount of $34,600, in accordance with the amendment fee schedule of 10 CFR 170.31.
If you have any questions regarding this matter, please write me at the above address or telephone me on 412/373-4652.
l Very truly yours, A. T.
Sabo, Director Licensing, Safeguards & Safety ATS/dc 8104010488
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LICBiSE C0fDITICIS Proposal For Improvement And Upgrading Of Operations at the Westinghouse Columbia fiuclear Fuel Fabrication Plant Sf;M-1107 Docket 70-1151 January 9,1981 e
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TABLE OF CONTENTS Page 201 Minimum Specificatiors and Capabilities 201 2.1 Scope of Licensed Activities 2 01 2.1.1
' Definitions 202.2 2.1.2 Summary Description 204 2.1.3 Material 205 2.2 Facility and Equipment Design features 205 2.2.1 General Requirements 206 2.2.2 Emergency Equipment 206 2.2.3 Personnel Monitoring Devices 207 2.2.4 Radiation Protection Measuring Instruments 207 2.2.5 Ventilation Equipment 209 2.2.6 Air Sampling Equipment 210 2.2.7 Low Level Liquid Radioactive Waste Disposal 210 2.2.8 Solid Radioactive Waste Disposal 210 2.2.9' Imr.ediate Evacuation Signal System 2.2.10 Processing Equipment Incorporation Nuclear 211 Criticality Safety Controls 2.2.11. Chemical Equipment 212 2.2.12 Incinerator Equipment 213 2.2.13 Moderation ' Control Areas 213.1
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2.2.14 Interlocks 213.1 2.3 Safety Limits 213.1 2.3.1 Chemical Reaction Safety 214 2.3.2 Nuclear Criticality Safety 2.3.2.1 General Nuclear Criticality Control 214 Procedures and Criteria 215 2.3.2.2 Nuclear Criticality Safety Values 2.3.2.3 Radiation Protection and Nuclear 234 Criticality Safety Administration 235 2.4 Minimum Conditions of Operation
'235 2.5. Emergency Procedures REVIstCn No.12 DATE:1/9/81 PAGE 200 DOCKET 70-1151 DATE:
8/24/74
Sffr1107 Minimum Specifications and Capabilities 2.1 Scope of Licensed Activities 2.1.1 Definitions Clean Area - an area where radioactive material, if present, is completely contained and there is negligible contamination on the floors or accessible surfaces.
(e.g. Machining Area, Grid Assembly Area, Final Assembly Area, Office Areas, Cafeteria, etc.)
Component - when used administratively, an organization unit, dist'inguishable by its assigned responsibilities, e.g.,
the line management component, the radiation protection component, etc.
Controlled Area - an area where uncontained radioactive materials are processed and probability of contamination on floors and accessible surfaces is high.
Protective clothing is required.
(e.g., Conversion Bay,etc.)
Area, Pelleting Area, Rod Loading Area, UF6 Dry - When used to describe special nuclear material (SflM), having a moderation ratio (H/U = hydrogen to uranium atomic ratio) less than, t
or equal to, 0.5 for uranium enriched up to 4.15 weight percent in U-235, and less than or equal to 0.3 for uranium enriched between 4.15 and 5.0 weight percent.
Equivalent Diameter - when evaluating the adequacy, for purposes of nuclear criticality safety, of the geometry control of a subcrit unit having a non-circular cross section, the diameter of that circle that has the same area as the area of the cross section of the subcrit dequiv.=kArea, where Area is the cross sectional area of unit.
the subcrit unit under review.
hsDM DATB 8/24/74 REVISION No. 12 DATE: 1/9/81 PAGE 201
5n-1107 l
2.1.2. Summary _?_escription The objective of the licensed activity will be the ADU or IDP, l
process corsersion of uranium hexafluoride to uranium dioxide and the manufacture of fuel-bearing components for nuclear The licensed material will be composed of reactor cores.
.unirradiated special nuclear materials received principally as uraniur.: hexafluoride containing uranium enriched up to 5.0 w/o l
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8/24/74' REVISION No.: 12 DATE: '1/9/81
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SM-1107 2.1.2 (continued)
Other chemical forms that may be received are uranium oxides and uranyl nitrate.
Operations may include the fabrication of fuel assemblies that contain uranium enriched from 4.15 to 5 w/o in the 235 The material for such assemblies will be processed U isotope.
through IDR conversion or will be received as uranium dioxide powder (only), and, will not be subject to ADU conversion, scrap preparation, or other wet chemical processes (unless properly approved and documented evaluations demonstrate the nuclear criticality safety of the utilized wet processing systems.) Another similar, limited activity will involve the fabrication of fuel assemblies containing mixed oxide fuel material. This material will be a mixture of Pu0 in natural or depleted U0.
It will be received and handled 2
2 only as sealed fuel rods that have beea fabricated by another Westinghouse flFD facility.
Such rods will be surveyed on receipt on a sample basis to '-he limits specified for sealed Pu sources.
flo authorization to possess or use exposed plutonium is intended.
Scrap or waste licensed material (< 4.15 w/o 235 ) resulting from l
0 site operations may be processed for concentration, recovery and/or These operations may involve chemical separation (e.g.,
reuse.
acid treatment and dissolution or acid leaching, followed by chemical precipitation), mechanical separation and themal decomposition.
Scrap and wastes resulting from processing operations that involve uranium enriched >4.15 w/o will be collected and segregated to assure that no materials at such higher enrichment are introduced into any area unless prcperly approved and documented evaluations demonstrate the nuclear criticality safety of the subject area at the higher enrichments.
D0cyrr70-115L DATE:
2/28/74 REVISION No.12 DATE:
1/9/81 PAGE 203 '
S W 1107 2.1.2 (continued)
The licenset
-tivity will perform work for other Westinghouse Divisions or outside customers which is adapted to the capabilities of the facility.
The work may consist of uranium oxide fuel fabrication and quality assurance testing operations and laboratory analyses of uranium or byproduct material.
2.1.3 Material Special Nuclear Matertal Listed below are the maximum quantities of special nuclear materials which will be possessed by the licensed activity at any one time.
Material Form
_ Quantity 235 Unirradiated, any chemical or 350 grams 0
or physical form at any enrichment 235 Unirradiated, any chemical or 50,000 physical form at any enrichment kilograms g
5 5.00 w/o 233 Any chemical or physical form, 5 grams 0
for laboratory uses only 238 Sealed Sources 1.5 grams Pu Mixed Unirradiated plutonium oxides, 750 kilograms Oxides mixed with oxides of natural or -
contained Pu.
depleted uranium, as sealed fuel Natural or rods.
The~ fissile Pu02 depleted U to
[(239Pu + 241Pu)0 ] will con-suit.
2 stitute a maximum of 6.6 w/o of the total oxide weight.
6 DDCKEr 70-1151 DATE:.
2/28/74 REVISION No. 12 DATE:l/9/81 PAGE 204
SfM-1107 2.2.11 Chemical Eguipment The equipment specified below will be provided as part of the UF6 vaporization system.
A UF detection method in the steam condensate system, with an 6
alarm to alert operating personnel to a leak in steam-type vaporizers, and with an interlock and an alarm to alert operating personnel to a leak in hot-water-type vaporizers.
A pressure relief valve and a liquid level detector in the steam-type UF vaporizers; or, with favorable geometry / overflow-type sumps in 6
the hot-water-type UF vaporizers.
6 Provisions to permit the leak testing of the UF -cylinder-to-conversion-6 system connections prior to heating each time a cylinder is connected.
A means for cooling the UF cylinder.
6 A means to prevent the backflow of water from the hydrolysis tank to the UF cyl.nder, in ADU process systems.
l 6
2.2.12 Incinerator Equipment The equipment, controls and safety interlocks specified below will be provided as part of the incinerator system:
Temperature sensor / controllers in the primary combustion chamber, breech and scrubber exhaust.
Safety interlocks to inhibit feeding additional wastes if an overtemperature occurs in the primary combustion chamber.
A means for automatic shut - down of the incinerator in case of overtemperature in the scrubber exhaust.
An auxiliary means for cooling the incinerator exhaust gases.
Docxtr 70-1151 DATE:
8/28/74.
REVISION No.12 DATE: 1/9/81 Paas 212
Si&l107 b
2.2.12
{ continued)
Cooling water flow monitoring devices.
Pressure monitoring devices in the breech and scrubber exhaust A means for automatic shutdown of the incinerator in case of insufficient negative pressure in the breech and scrubber exhaust.
Continuous, representative gaseous effluent sampler.
HEPA filtered exhaust and ash removal systems.
A means for monitoring and adjustment of the pH of scrubber solutions.
235 Maintenance of a log indicating the mass of 0 charged and removed for each burn cycle, and the cumulative total of the net, assumed to remain in the incinerator.
2.2.13 hoderation Control Areas Fire. control in areas of the SNM Building where uranium is processed, h'andled, or stored under hydrogeneous material criticality control criteria shall receive particular attention, as follows:
Special consideration shall be given to use of fire-resistive or noncombustible building components, equipment, and materials.
Special consideration shall be given to the prompt disposal of combustible waste.
Such weste that is collected during work activities shall be stored in metal containers having fire protection covers.
A readily available supply of portable fire extinguishers suitable for use on the specific hazards encountered shall be provided.
Such areas shall be subject to administratise controls, including specific personnel training, to assure that only permissible firefighting means and materials are used.
Revision No.12 DATE:
1/9/81 Pace 213 DOCKET 70-1151 DATE:
8/28/74
sp;4-1107 4
2.2.14 Interlocks IDR and moderation controlled blending and storage equipment, and its associated control instrumentation, shall be evaluated for conditions requiring automatically operating interlocks to safeguard facilities, workers, and environs against failures of equipment and instruments signficant to safety.
Equipment and instruments requiring such interlocks shall be identified
[ and documented by the regulatory compliance component.
Identified interlocks shall be installed, maintained, and operable as a minimum condition of applicable equipment or system operation.
.3 Safety Limits 2.3.1 Chemical Reaction Safety Each anion-type ion exchange column will be equipped with a rupture disc.
The concentration of the nitric acid regenerant will be controlled at or below 5.0 normal. The columns will be maintained under rou'.ine surveillance during regeneration.
When the ion exchange columns are not in routine use, the bottom of the column shall be opened to atmosphere by removing the spool piece.
9 DOCKER 70-1L%
DATE:
8/28/74 REVISION No.12 DATE:
1/9/81 PAGE 213.l
4 Stfi-1107
'2.3.2.2 Nuclear Critical.ty Safety Values Maximum Permissible Values Maximum Permissible Values for subcrits with a maximum a.
2350 enrichment of 5.0 w/o are established in tabular form as follows:
Figure 2.3.2.1 Batch or Mass Controlled Subcrits Figure 2.3.2.2 Volume Controlled Subcrits Figure 2.3.2.3 Cylinder Diameter or Equal Cross Sectional Area Controlled Subcrits Figure 2.3.2.4 Slab Thickness Controlled Subcrits b.
Subcrits containing uranium enriched to greater than 5.0 w/o will be. limited to 350 grams of contained 235U.
Moderation Controlled Subcrits - Systems shall be considere'd c.
under moderation control for nuclear criticality safety when the following conditions are met:
The contained special nuclear material is " dry" unde: normal operating conditions.
The containment precludes introduction of moderator; or, system controls, procedures, and interlocks preclude intro-duction of sufficient moderator to compromise the nuclear criticality safety of the system.
in approved shipping cylinders Moderation controlled UF6 will constitute a specific MPV.
d.
Uranium concentration controlled subcrits will be limited 235U to a maximum, allowable concentration of 5 grams per liter. This MPV will not be applied unless it can be demonstrated that the precipitation of the SNM and higher concentrations due to process failures are not credible.
DOCKER 70-1151
.DATE: 8/24/74 REVISION Ib.12 DATE: 1/9/81 PAGE 215
Sffi-1107 2.3.2.2 (continued)
Subcrits which are safe by concentration control and e.
which are part of a continuous processing line, will be filled with Raschig rings which are maintained in accordance with the current edition of the standard N16.4, "Use of Borosilicate - Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material."
f.
. Fixed poisons may be used in nonliquid special nuclear material systems when the following restrictions are both met:
The poison shall be physically protected from abrasive action by the special nuclear material; and, Nuclear Criticality safety of such poisoned systems shall be verified by validated computer calculations.
Subcrits composed of fuel assemblies will be limited l
g.
by reactivity. The computed keff, including allowances for computational error, will not exceed 0.95.
These computations trill be performed by the -
NFD, Nuclear Design Department using procedures such as MUFT, SOFOCAT, LEOPARD, and/or PDQ-03.
The' results of these computations will be independently reviewed within the department and approved by the department manager before being transmitted to the NFD Manufacturing Department.
h.
Maximum permissible values for subcrits containing l
plutonium will be those established in Figure 2.3.2.5.
Subcrits containing plutonium will be restricted mixed to encapsulated components containing Pu02 with natural or depleted UO '
2 DOCKET 70.J13 DATE:
8/28/74 REVISION No.12 DATE: '1/9/81 PAGE220
SiYr-1107 2.3.2.2 (continued) d.
Moderation controlled subcrits (" dry"*) of 5 w/o l
or less enriched uranium, when limited as follows will not be considered to contribute to interacting arrays:
(1) Maximum Permissible Values, (2) in closed containers or configurations which would not retain water, (3) located outside of areas assigned to interacting subcrits, (4) no sprinkler system in the area, (5) no use of water or other hydrogenous agents for fire fighting purposes and (6) appropriate nuclear criticality safety signs posted in the controlled area.
Concentration controlled subcrits (with or without e.
borosilicate glass Raschig rings) are not considered to contribute to nuclear interaction provided that they.are outside of areas assigned to interacting subcrits.
f.
Notwithstanding other spacing requirements, any subcrit will be separated by at least 12 inches from any other subcrit.
All subcrits containing plutonium will be spaced g.
such that the " smeared" slab thickness will not exceed 25% of the minimum critical slab thickness.
l
- As defined in paragraph 2.1.1
! DOCKET 70-llM DATE: 8/28p4 -
REVISION No. 12 DATE:.1/9/81 PAGE 232 '
Sil4-1107 2.4 Minimum Conditions of Operation Processing operations involving SflM will be performed routinely only when appropriate equipment having the capabilities specified in paragraphs 2.2.3 through 2.2.14 has been provided and is operative l
and when qualified line management personnel are present.
Non-routine and emergency operations involving Stim will be performed only after the particular operation has been approved by the appropriate line management function and the radiation protection function. The line manager will be responsible to obtain the
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evaluation and approval of the operation by the radiation protection function, consistent with the urgency of the situation in the event of an emergency.
All equipment specified as necessary to the operation by the line management function and radiation protection function will be provided and operative.
The equipment specified in paragraph 2.2.2 will be maintained available a't all tii..es.
The necessary trained personnel will be available as specified in the emergency procedures.
A continuing program of surveillance, air sampling and smear sampling which will be adquate to detect ventilation deficiencies and assure compliance with 10 CFR 20 limits, will be conducted.
2.5 Emergency procedures Written emergency procedures which comply with the requirements of 10 CFR 70.22(i) will be maintained and communicated to all employees and unescorted personnel working'in the affected areas of the licensed activity.
Selected personnel will be organized and trained to cope with various credible emergency situations.
DOCKET 70-1151 DATE: 8/28/74 REVISION No.12 DATE: 1/9/81 PAGE'235
e g,,
TABLE OF CONTENTS 276 4.
Authorizations and Exemptions 276 4.1 Authorization for Use of Materials at Off-Site Locations 277 4.2 Abandonment or Disposition of Materials or Equipment 277,'
4.3 Record Storage 278 4.4 Exemptions from the Requirements of 10CFR70.24 278 4.4.1 Isolated Areas 279 4.4.2 Low Concentration Storage Areas 279 4.4.3 Shipping Package Storage Areas 280 4.5 Non-Radioactive Industrial Wastes 280 4.6 Possession of Licensed Materials at Reactor Sites l
282 4.7 Transfer of Special Nuclear Material l-i t
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g g.:70- M DATE:- 8/28/74 REVISION No.12 DATE: 1/9/81 pK2E275 '
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SIM107 Transfer of Special fluclear flaterial 4,7 _
The licensed activity shall be authorized to sell, or otherwise transfer, hydrofluoric acid containing trace quantities of uranium (enriched up to five weight percent in U-235) to nonlicensed persons, provided that:
The concentration of uranium in the acid does not exceed 10 parts per million.
Each such sale or transfer shall be accompanied by a written instruction that the acid is not to be used for any purpose involving human consumption.
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DOCKET 70-1151 DATE: 8/28/74 REVISIONNo.12 DATE: 1/9/81.
PAGE282 l
BNIR0ffD(TAL INFORMTION Proposal For Improvement And Upgrading Of Operations at the Westinghouse Columbia Nuclear Fuel Fabrication Plant SNM-1107 Docket 70-1151 L
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January 9, 1981
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SECTION 1 DESCRIPTION OF PLANT CHANGES To best meet the established need for increased productivity at the Columbia Plant, a fully developed and proven dry conversion process - the Integrated Dry Route (IDR) method developed and commerciallj utilized by British Nuclear Fuels Limited (BNFL) - is proposed as a supplement to the plant's existing ADU (wet conversion) process production lines.
The planned IDR process line will replace an experimental dry process line -
the Direct Conversion Fluidized Bed (DCFB) system - which, although it has been shown to provide some of the desired environmental advantages of the IDR process, it has not provided the superior fuel product anticipated in the new process.
The IDR process line will provide opportunity for productivity improvement, while generating much lower quantities of regulated constituents in liquid effluents, and, while also providing enhanced control of regulated airborne constituents (when both are compared to equivalent ADU fuel production capacity).
Details of the proposed IDR system, and plant changes to accommodate installation of the total Manufacturing Automation Project (MAP), are described in the following discussion.
1.1 OVERALL PLANT OPERATIONS The fabrication of nuclear fuel assenblies requires both chemical and mechanical operations; and, as a result, some (low level radioactive)
. solid, liquid, and' gaseous wastes are generated. The plant process equipment and' ventilation systems are designed and operated to maintain regulated chemical and radioactivity discharges to the environment well
~
within established limits,.and as low as reasonably achievable.
Figure 1 presents a general flow diagram of proposed total plant operations, including the new process line.
(This figure also schematically identifies the sources, treatment,1and anticipated effluent concentrations in releases to the environs, from new system-installation).. Figure 2 presents a plan view and Figure 3 shows a cross-sectional view of the proposed process area.
1.2 SYSTEM DESCRIPTION 1-1
1.2.1 Conversion The planned conversion process will utilize dry methods to convert solid uranium hexafluoride (UF ) to uranium dioxide (U0 ) powder.
[This process 6
2 involves UF vaporization, gas phase hy.olysis, and gas-solid phase 6
reduction to produce U0. The process is well established, has been 2
commercially utilized in two countries, and provides opportunity for significant environmental advantages over alternate processes (by substanitally reducing liquid waste generation).]
UF feed material, received in type 30A/30B cylinders, is vaporized within 6
the cylinders by heating with hot water spray. The resulting UF vapor 6
is reacted with superheated steam at the head-end of a conversion kiln to The form uranyl fluoride (U0 F ) p wder and hydrogen fluoride (HF) gas.
22 U0 F is further contacted within the kiln - with a countercurrent flow of 22 hydrogen, nitrogen, and superheated steam - to strip residual fluoride, and to reduce-the uranium powder to uranium dioxide. The U0 is discharged 2
from the product end of the kiln into check hoppers, and is then pneumatically Process conveyed (or otherwise transported) to the. powder processing area.
off-gases { hydrogen (H ), hydrogen fluoride (HF), nitrogen (N ), and steam 2
2
. H 0)] are removed continuously from the top of the head-end of the kiln, (2
through process filtration (pariodically reverse-purged) for retention of uranium-bearing solids prior to recovery of hydrofluoric acid.
The conversion process is shown schematically in Figure 4; details of the
~
conversion process follow:
Vaporization' cylinders are transported from the UF. storage area to the Manufacturing UF 6
6 Building UF bay via lift truck.- The cylinders are then installed in a vaporizer, 6
using an overhead crane. A cylinder is connected.to its process header by flexible copper tubing. When the cylinder under hot' water spray has reached is process operating temperature (% 180*F) and pressure (> 5 psig), the UF6 delivered'to the process (kiln) by remotely opening the cylinder valve. 2 9
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[Normally (during steady-state operations), two cylinders will be hot simultaneously, but during changeover periods (some 50 percent of the time),
three cylinders will be hot simultaneously.]
When 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 disconnected from the supply line and valved into a cold trap evacuation system for removal of residual UF6 (to an acceptable final heel of less than 25 pounds).
Heel removal is accomplished by evacuating the cylinder (with a vacuum pump) through an exhaust train, consisting of a cold trap system with vapor) and two final self-contained (-65'F) refrigeration (to condense UF6 series chemical absorber (A10 ) traps (for the capture of any final traces 23 ofUF).
Upon completion of the evacuation process, the cylinder is removed 6
from the vaporizer, and transferred by crane to the cylinoer scales for weighing to assure that the residual heel is equal to (or less than) 25 pounds.
vapor or liquid containing lines and vessels are In the vaporization area, UF6 normally enclosed within a pipe chase (or other containment), maintained at a negative pressure with respect to the operating area and vented through the HF Vent Scrubber.
UO Powder Production and Handling 2
Powder product in an IDR kiln.
UF supplied by vaporization is converted to UO2 6
Converted powder product from a kiln is held in discharge check hoppers pending samples are used to analysis and subsequent processing. Composited U02 establish the physical corposition of the contents of each check hopper.
Acceptable product powder is discharged to a pneumatic transfer.line which conveys the material directly to powder preparation.
(Bulk powder centainers are also to be provided as an alternate transport system.) Powder in a check hopper found'to be unacceptable (with respect to fluoride and/or moisture) is transferred to a powder rework area for further treatment.
Powder Rework Rework is necessary when powder properties (primarily fluoride and moisture content) are out of specification. Powder which does not meet specifications 1-3
is discharged from the check hoppers into (geometrically-controlled) containers, then is campaign processed through auxiliary drying and fluoride stripping equipment.
Reworked powder, which upon analysis is found to meet specifications, is returned to the regular process stream by one of two methods: When the pneumatic conveying system is used, a pre-wieghed amount of reworked powder will be metered into a transfer line af ter a receiving blender has been charged with the designated amount of virgin powder. When the bulk transport containers are used, the reworked powder will be accumulated in a (moderation-controlled) container until it is full. The container will then be elevated to the blender charging floor through the container lift and used as needed to supply addback to each blender charge.
Hydmfiuoric Acid Recovery The conversion kiln off-gas is cooled to recover byproduct hydrofluoric acid by condensation. The recovered acid solution is collected in an HF quarantine tank (Q-Tank) and held for uranium analysis before release (if it meets uranium specification) to a bulk storage tank (for subsequent sale as a byproduct).
Out-of-specification' (excess uranium) HF solution is transferred to a safe-geometry precipitator system, located in the acid recovery area, where uranium values are-recovered by (advanced treatment) precipitation and filtration.
The filtrate is transferred to the liquid waste treatment system.
Based on the current HF condenser design, some 95% of the HF will be recovered as (nominally) 55 w/o hydrofluoric acid. The remainder of the HF will be cleaned by Na0H in the Condenser Off-Gas Scrubber. The annual HF transfer to the scrubber liquor is estimated at some 4400 Kg (approximately 10,000 pounds).
[It is planned:that the recovered HF. be licensed for recycle (e.g., sale);
otherwise, it would have to be neutralized with Ca(OH)2 with the product (CaF ) dried and buried.
Licensing for recycle will allow the reuse of the 2
recovered acid in an economic manner which would also conserve resources -
.rather than requiring the otherwise unnecessary controlled disposal of the material, as radioactive waste, in valuable space at a licensed burial ' site.]
1-4' e
Off-Gas Scrubbing from gaseous exhaust Caustic scrubbing is provided to remcve residual HF and UF6 streams - from either the nomal process (and vent exhausts) of the acid cylinders or transfer lines.
recovery system, or from an inadvertent leak in UF6 Two basic scrubber systems are provided for this purpose:
The Condenser-0ff-Gas Scrubber is provided to cleanse condenser effluent from the kiln system (which contains traces of hydrogen in addition to acidic off-gas).
Uncondensed HF gas carryover is treated with caustic (Na0H) in this The resultant scrubber solution (containing traces of uranium) is scrubber.
quarantined and analyzed prior to release to liquid waste treatment, or to the Residual precipitation tanks (if established uranium levels are exceeded).
hydrogen (already below the lower limt of flammability) is further diluted with plant air prior to discharge of the gaseous stream to the outside atmosphere (through the confinement and ventilation system de;cribed in paragraph 1.3).
The HF Vent Scrubber is utilized to cleanse the nomal off-gas from vents on HF storage vessels.
In addition, this scrubber can be made available to cleanse air in the vaporizer room, and vaporizer room process enclosures, which might Process piping or vessels.
become necessary due to inadvertent leakage of UF6 HF Vent Scrubber solutions and gaseous streams are routed in a similar manner to that previously described for Condenser Off-Gas Scrubbing, except that there is no need for hydrogen treatnent.
Uranium Recovery from Solutions Uranium will be recovered from spent scrubber solutions (as well as from uranium-containing by-product HF filtrate) via Columbia's existing Advanced
. Waste Treatment system.
1.2.2
-Fabrication Details of the fabrication process follow:
Powder Processing and Pellet Fabrication 2 p wder from the dry conversion process is transferred to the powder proces UO ing area where it. is blended with additives - including uranium oxide (U 0 )
38
.1-5
4 recycled from scrap recovery processes. After blending, the homogenized powder is compacted, granulated and pressed into pellets.
There are no liquid effluents from these operations; airborne effluents are treated by the confinement and ventilation system described in paragraph 1.3.
Sintering Pressed pellets are loaded into cats and charged to electrically heated furnaces for transformation to high-density pellets by sintering in a reducing atmosphere.
There are no liquid effluents from this operation; airborne effluents are treated by the described confinement and ventilation system.
Pellet Grinding and Rod Loading Sintered pellets are processed through a grinding operation to obtain specified dimensions.
Ground pellets are loaded into prepared metal tubes (from the Tube Prep Area) and the tubes are sealed by welding.
Finished rods are inspected and tested, then transferred for final assembly. There are no liquid effluents from these operations; airborne effluents are treated by the described confinement and ventilation system.
Final Assembly In the IDR portion of the' Final Fuel Assembly Area, the fuel rods are loaded into designeted positions in a prefabricated support' structure consisting of a bottom nozzle, thimble tubes, and structural grids. A top nozzle is then attached to complete the final assenbly. There are no liquid or airborne effluents from these operations.
1.3 Confinement and Process Ventilation As noted ~in the introduction to th'is Section (and reiterated in several other
~
paragraphs), the dry process provides the opportunity of enhanced control of airborne effluents, through improvements in containment and processing.
The MAP Confinement land Ventilation System (within the controlled area of the Manufacturing Building). functions to enhance limitation of plant personnel
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exposure potential, and to enhance protection of the general public by strict control of airborne effluents discharged to the environment.
A generic ventilation schematic is depicted in Figure 5.
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SECTIONAL VIEW MAP PROCESS AREAS
- WATER BARRIER FOR MODERATION CONTROL HF REC 0VERY PNEUMATIC TRANSPORT SYSTEM FILTER ROOM BLENDING
[ MECHANICAL NEW PORTION EQUIPMENT r i r e / / AREA /r' gf CONVERSION / -[ AREA / l -,9 r. 1 FINAL I FUEL R0D FlBRICATIOb FUEL ASSEMBLY _'( \\ TUBE UT FURNACE GRINDING & l AREA AREA INSPECTION & INSPECTION PREP WORIZATION fi l M /H I A l COMPACTION & PELLETING U CONTROL gygg R00 & DISCHARGE CONTAINER pgp SCRAP LIFT REC 0VERY AREA
- SEE FIGURE 2 FOR SECTION LOCATION FIGURE 3
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e EXHAUST TO ENVIRONMENT AL r X FINAL HEPA FILTER INITI A L ROOM AIR HEPA ir NON-(LEVEL 1) CHEMICAL CHEMICAL CON FINE. CONF INE ME NT SCRUBBER MENT ILEVEL 11 4L (LEVE L 21 .tk PRESSURE GRADIENT = P <P2<E3 g ' (LEVEL 3) si SUPPLY AIR 1 L Confinement an' Ventilation ~ Schematic for NAP Area d Figure 5 - 12
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SECTION 2 ENVIRONMENTAL ANALYSIS The proposed installation of an Integrated Dry Route (IDR) process line at the Nuclear Fuel Division's Columbia Plant requires minor modifications to the existing licensed facility and will result in minor incremental releases of radioactivity and chemicals to the environment. However, expected emissions are much less than those estimated in the 1975 Environmental Evaluation, less than the existing.ADU emissions, and less than an equivalent addition of ADU capacity. The purpose of this Environmental Analysis is to compare these effects with those previously approved by the NRC in the 1975 Environmental Evaluation. A. Area Preparation 1. gxistjpg,ggufpgegt_Reggya! The proposed IDR line will replace the existing Direct Conversion Fluidized Bed (DCFB) line and the obsolescent incinerator which are described in the existing license. The disassembly, decontamination and dispositioning of the DCFB line and obsolescent incinerator are part of an on going program at the Columbia Plant to remove and replace non-productive equipment. Consequently, these plans had been developed'and are being implemented independent of the proposed IDR installation. Presently, instrumentation is being renoved for possible reuse in other areas of the facility. Future efforts will involve dissassembly and dispositioning of the remaining portions of the DCFB line and obsolescent incinerator. ' Dismantling existing equipment to make room for the IDR system is
- being performed according to existing license conditions.
Existing . procedures and administrative controls are' being used to contain airborne radioactivity and minimize traffic through the affected For example, where. necessary, temporary enclosures will areas. be constructed to minimize releases of airborne uranium; and a negative pressure.(below atmospheric) ventilation and exhaust system is being maintained in the construction area. 2-1 't
e Previous similar experience with equipment removal, decontamination and dispositioning has shown that these activities can be performed with negligible impact on the environment. Examples include: (1) replacement of the pellet line feed ends with improved equipment to facilitate uranium processing and minimize airborne radioactivity, (2) removal of obsolete exhaust ventilation equipment on the facility roof and replacement with state-of-the-art equipment, and (3) construction of an approximately 100,000 square foot addition to the Columbia Plant. 2. Bujjpjpg,$ggjfjgatjggs The proposed line will be installed between columns Al-B1 through A9-B9 of the existing Manufacturing Building (as shown in Figure 2). Changes in plant facilities will include elevating the roof in the area between columns Al-Cl to A6-C6 (Figure 2), and modifying existing structures to support the additional load - as required by the equipment and the roof superstructure. The facility changes required for the installation of the MAP system include: (1) removing part of the concrete floor in the area designated for the IDR line (Figure 2) - as necessary to provide the additional structural support for equipment and the roof super-structure, (2) extending the roof line upward approximately 20-30 feet above existing roof line in the designated area, (3) installing structural steel necessary for IDR equipment and building support, (4) modifying the heating, ventilation and air conditioning (HVAC) system located in the steel trusses in the ceiling above the installation area, (5) replacing filter-exhaust systems located on the piping and conduit systems within the installation area to meet moderation water barrier requirements. Inside the designated area, some of the concrete floor will be ~ broken up and removed (and some dirt underenath might be removed) to permit installation of support structures. Exhaust filter systems located on the existing roof of the designated area will be scrapped according to existing licensed procedures. Equipment'to be removed will include the DCFB exhaust, the emergency UF -DCFB' exhaust, and the obsolescent incinerator exhaust, Roof 6 2-2 L
penetrations will be temporarily covered during construction so that there will be no release of airborne constituents, nor leakage of rain or snow into the production areas. The required openings will be promptly covered and sealed in a permanent manner. Continuous air monitoring of existing roof exhaust stacks will be conducted; negative pressure within the manufacturing building will be maintained at all times. Penetrations through the existing roof will also be required for the a new roof sup rstructure installation, and for new equipment installation. These temporary open penetrations will be handled the same as those required for removal of existing equipment. 3. Rel oc a t i gn_o f, Ex i s t i ng,P l a n t_ S erv i c e s Certain plant services within the planned installation area will have to be relocated. These include the Health Physics Laboratory on the operating floor, and the Machine Shop, which will be moved to another area within the existing building. 4. Summary For the proposed activity, there will be no significant construction impact. The floor area affected by the IDR system installation will amount to approximately 22,000 square feet, or only about six percent of the existing manufacturing building floor area; and, the roof superstructure will also enclose approximately 22,000 square feet, or only about six percent of the existing plant roof area. Thus the incremental plant area which will be temporarily affected by the dismantling, construction and installation activities is a relatively minor percerJ.r.ge of the total plant area, and planned activities would most certainly be expected to cause much less effect than the approximately 30 percent floor area addition accomplished in 1978 (Amendment #2 to SNM-1107). B. Effluents Minor effects on tha environment resulting from normal operations might be expected to occur as a result of the addition of the automated IDR 2-3
= _ - process line. There are, however, no changes in the types of effluents from the IDR process when compared to the existing ADU lines. The magnitude of such plant radiological and chemical impacts are evaluated in this section. The pathways for potential dispeision of radioactive or chemical discharges to the environment are the same for the new IDR line as for the previously evalnated ADU and DCFB processes, differing only i in relative magnitude for the respective airborne and liquidborne releases. Thus, there are no unforeseen or unevaluated effects introduced by addition of the IDR line. The relatively small effects for liquid and air waste discharges are evaluated below. Table 1 shows the airborne and liquidborne releases from the proposed IDR line and compares them with the 1975 Environmental Evaluation, recent ADU performance and effluents from the existing DCFB line and obsolescent incinerator. This Table shows that: (1) proposed IDR releases are much less than t%se previously evaluated in the 1975 Environmental Evaluation, (2) except for airborne fluoride releases, IDR effluents are expected to be below the release of the combined ~ DCFB and obsolescent incinerator which are being replaced by the IDR system (the expected airborne fluoride levels are well below the 1975 Environmental Evaluation estimates) and (3) except for airborne fluoride releases, IDR effluents are expected to be below existing ADU performance levels. All IDR liquid effluents are expected to be well'within EPA limits for the existing Columbia Plant National Pollutant Discharge Elimination System Permit (NPDES). - Consequently, no changes in this permit will be required. Estimated stack release rates of. uranium and fluoride in micrograms ~ per second are summarized below: 2-4 ~
9 Fluoride Uranium Micrograms /sec Micrograms /sec Previously Estimated (1600 MTU/ year) 24,000 75 Existing ADU (700 MTU/ year) 660 31 Estimated IDR (500 MTU/ year) 2,125 3.5 These data (and Table 1 data) show that fluoride emissions will be con-siderably less than those previously estimated but somewhat greater than existing ADU emissions. Estimated airborne fluoride concentrations at the site boundary, however, are much less than the most restrictive State limit for fluorides (0.5 micrograms per cubic meter). During initial operation of the IDR line, representative state-of-the-art stack sampling of fluorides will be performed to verify the expected low effluent concentrations. Note that the existing ADU fluoride emissions are much less than the 1975 previous estimates. This is attributed to the fact that fluoride scrubbing converts the fluoride to a particulate form which is collected with a high efficiency by the following stage of HEPA filtration. A similar effect is expected with the IDR Scrubber /HEPA filtration system, with actual effluents being less than those estimated. C. Production Throughput The 1975 Environmental Evaluation estimated environmental impacts based upon an annual uranium thrcughput of 1600 MTU. The expected throughput of the Columbia Plant with the addition of the IDR process will continue to be less than 1600 MTU per year. D. Effects of Accidents The Westinghouse Columbia Plant License documents have previously evaluated a spectrum of hypothetical plant accidents (ranging in severity of consequences from trivial to significant, and ranging in probability of occurrence from credible to incredible). In considering the addition of an automated IDR process line at the. Columbia plant, this spectrum of accidents was reviewed, and an evaluation has been performed for additional postulated accidents which could potentially occur.as a result of proposed plant improvements. 2-5
Sets of hypothetical accidents for the IDR line shows that (1) the types and severity of postulated accidents are similar to those evaluated in the 1975 Environmental Evaluation, (2) the effects of hypothetical accidents are well within those evaluated in the 1975 Environmental Evaluation, and (3) the probability of occurrence and severity of Category 3 accidents (Maximum Credible Accidents) as defined in the 1975 Environmental Evaluation are not changed with the addition of the IDR line. E. Conclusion The proposed installation of the IDR process will result in a significantly improved method of uranium conversion and fabrication, when compated with an' equivalent addition of ADU capacity, while guaranteeing minimal environmental-impact. The following benefits will be achieved with the IDR process: 1. Increased automation will result in fewer individuals . required to operate the lines and thus lower occ:1pational exposures. 2. The IDR process represents an improvement in process control and containment when compared with the ADU process which should result in reduced in-plant airborne radioactivity concentrations and thus reduced personnel exposures. 3. The reduction in the aqueous fluoride emissions will minimize the solid calcium fluoride wastes now generated in the ADU process. 4. Ammonia releases are not a factor'as they are in the i ADU process since ammonia is not used in the IDR process. i '5. The removal of-waste. fluorides as hydrofluoric acid will result in the potential recycle of this material. 2-6 9 9
TABLE 1 Comparison of Airborne and Liouldborne Constituent Releases from Proposed IDR Line, with 1979-1980 ADU Performatice,. .and Previously Evaluated Releases Type-of. Effluent (1) That May Be 1975 Westinghouse (1600 MTU) .IDR DCFB + 01 1979-80 ADV Media Released Offsite Environmental Evaluation Increment Increment. Performance Units Liquid . Uranium (U) .l.500 0.020 0.023 0.91 Pounds Per Day Ai r'
- U.
0.00096 0.00005 0.00005 0.0004 Micrograms Per Cubic Meters (2) m - Liquid Fluoride (F) 25.0 0.5 0.1 19.6 Pounds Per Day 4 Air F 0.3100 0.0275 0.0110 0.0084 Micrograms ~ (DCFB only) Per Cubig Meters (2i -Liquid Total Suspended 25.0 0.5 0.2 11.7 Pounds Per Solids Day l Liquid pli 6.0-9.0 8.5 7.0 8.9 pil Units 1 ~ L(1). DCFB.Line.plus Obsoles' cent Incinerator j (2) Calculated concentrations at the site boundary (1800 ft NNW of the plant) L.}}