ML19281B325
| ML19281B325 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 04/03/1979 |
| From: | Dipiazza R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML19281B327 | List: |
| References | |
| WRD-LS&S-703, NUDOCS 7905090190 | |
| Download: ML19281B325 (53) | |
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WeNinghouse Water Reactor Ec h N[pp{yop,ogg ons 2 40 WRD-LS&S-703 o
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U.
S.
Nuclear Regulatory Commission Office of Nuclear Material"'sa~fety & Safeguards A
Division of Fuel Cycle & Material Safety m
Washington, D.
C.
20555 Attention:
Mr.
L. C. Rouse, Chief J
Fuel Processing & Fabrication Branch ~;
Reference:
WRD-LS&S-683, dated March 7, 1979 Gentlemen:
Subject:
Retransmitta~ of Application for Amendment to Expand Facility, License SNM-1107, Docket 70-1151 The Westinghouse Electric Corporation hereby requests an amendment to License SNM-1107 to authorize operations with special nuclear material in the expansion to our Columbia Facility, in accordance with the attached application.
The material transmitted by the reference is to be discarded in its entirety and replaced with the attached material.
The proprietary portion of this application is being transmitted under separate correspondence in accordance with the provisions of 10 CFR 2.790.
Please find enclosed a check payable to the U. S. Nucler-Regulatory Commission in the amount of $34,600 in accordance with 10CFR170.31.
If you have any questions regarding this matter, please write to me at the above address or telephone me on 412-373-4652.
Very truly yours,
~
-- THIS DOCUMEilT CONTNUS k^
I' 4%$4 POOR QUALITY PAGES Gu /ld Ronald P. DiPiazza, Managet NES License Administration
&'lN AL slw/ Attachment i
j 7905090/70 g.
CDLUMBIA PLM7r EXPANSION This expansion consists of the constructica of an approximately 100,000 square feet addition to the present nunufacturing building.
75,000 square feet of the new areais n11 m ated for factory space, 5,000 square feet fcr truck docks and 20,000 square feet for nez~h. This addition is adequate to house nea pro-duct iwguvenent equipent, a cbmim1 process facility, equipent to increase production and increased in-process inventory. Also included for use in this area are a secoM incineration systen for recovery of uraniun fran waste naterials and a systen for uraniun recovery fran scrap by solvent e. Men.
'Ihe building expansion will be effected by t:he extension of the present manu-facturing space in the north and west directions, Figure 1.3.2.1. The exterior walls will be prefabricated, prestressed concrete, tee p els which are consistent with l
the stringent specifications of the present buildmg.
t Approx 1nntely 75,000 of the 100,000 square feet space will be used for new cr expanded ch mi m1 and mechanical areas. The remainire space will be utilized t
for in-plant offices ard service areas.
'Ihree basic new itens will be operated in the eW. chenical nanufacturing 4
area. Each is listed below with a brief description of the operation and/or equignent.
I (1)
Solvent Extraction i
'Ihe solvent extraction equipent includos a dissolver systen similar to the units ncw used in the present scrap recovery operation, preparatory equipent, feed and adjustrent tanks, and extraction column and a stripping column, storage and blending tanks, sludge concentration and' collection l
systans and necessary piping, ventilation, instru::entation, and operation i
platf6nn.
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(2)
Chenical PImess. Develogent Facility
'Ihe basic equiptent for the develcpunt lab will provide an environment for
'N ot&90-/ M.
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experimental work.
A walk-in ventilated hood, small fume hood and
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dust collectors will be used for dust and fune control.
A. tube calciner and sintering furnace units will be used to simulate the present production units for develognent purposes. Material handling, scales, etc. are ex-anples of support equignent to be utilized in the lab. Special equipmnt for specific developmnt projects will be provided as neMM.
(3)
Incinerator Svstan The second incinerator system will be installed in the new building nadition.
The systen will consist of an incinerator, quench tower, absorber tower, heat excharger, surp tank, condenser, reheater, oil tank, filter house and noter control center.
'Ihe expansion area will be covered by the plant criticality alarm system and all personnel working in this area will be covered by the li nsee's bioassay s u cm i
4 as described in Section 3.2.3 of the present license.
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PIRE EXPAN3 ION VS W aVIRON:EM. EVAUETION, BRPCH 1975 i
'Ibe proposed operations in the plant expansion have been ccupared with the March 1975 W Envirormental Evaluation.
In general, all parannters are within the limits based upon a throughput of 1600 nutric tons of uraniun per year as discussed in the repert. The follcwing table lists the Envircraental Evaluation ccmnitnunts and the corresponding plant expansion parameter:
HWIPOt@EtEAL EVARATION CQ4TISEE PIRE EXPANSION PAPNEIER 1.
h developed area occupies less 1.
The plant expansion consists of an 2
l than 5% of the total site area of area of approx.100,000 ft. The 1158 acres (i.e. 1158 x.05 =
total disturbed area (including the 8
57.9 acres) existing building, plant expansion, lagoons, waste treat:nent, parking lot, 1
etc.) l' less than 50 acres.
2 I
2.
A 50,000 ft buildirq is described 2.
The totalarea of the plgnt e.wien i
to a W ate fabrication of mac-is of apprcx.100,000 ft and will 2
i
' hined components plus a 50,000 ft a w W ate additional operations to expansion of the contanunation increase capacity.
controlled area.
I 3.
The plant aMitions will increase 3.
The capacity is e. W to be less production to 1600 M / year.
than 1200 MrU/ year.
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4.
The building is of modern design, 4.
The plant expansion is architecturally architecturally expressed by its simile to the existing building.
long simple lirns of rectangular interconnecting areas, ard designed to ccuplement its rural flat sur-roundings.
5.
The total gaseous discharge rate to 5.
The total e.W gaseous discharge 3
envircraent is less than 160,000 rate will be less than 160,000 ft / min.
1 6.
Gaseous effluent discharge rates 6.
N average weekly gaseous effluent
{
were estimated at 26.94 alpha discharge rate for 1978 was less than i
uCi/ week (400 MIU/yr.) and 71.94 50 alpha uCi/ week. This is well withi_
alpha uCi/ week (1600 MrU/yr.
the envircrauntal estimate. The plant expansion contribution is expected to d
be minimal.
7.
Gaseous effluent exhaust stacks 7.
All effluents will be discharged thrcq
{
are described.
existire stacks except for the new in-cinerator which will e.42ust through a separate stack. Pctivity effluents f-l the incinerator are e.W to be les._
than the present incinerator.
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DNIBOtNENIAL EVAUJATICN CDMIDENT PIAVf EXPANSION PAPATTER 8.
Exhaust effluents will be main-8.
Average stack effluents for 1978 were tained belcw MEC.
well within MPC. Exhaust concentrat-ions for the plant expansion ventila-tion will also be well belcw MPC.
9.
Liquid waste stream concentrations 9.
Additional liquid wastes will include nust be less than 3 x 10-5 alpha solvent extraction tank system scrubber uCi/ml before leaving the plant.
water and incinerator scrubber water.
No significant increases over present operations are expected. All liquid effluents will be directed into existir Piping.
- 10. Solid wastes are either treated to
- 10. Burial disposal of embustible waste j
recover uraniun or buried ar. an NBC will be reduced with the additioral licensed burial ground. Pysuxi-incirarator capacity. 'Ihe total nately 40-60 bales / day of ccrtusti-quantity of non-cmbustible contaminat ble waste and 4 bales / day of non-con-waste is expected to decrease because l
bustible contaminated waste will be of increased scrap recovery cap,hiliti j
gencrated.
}
- 11. Expcsure pathways include uraniun
- 11. No new pathrays are expected.
. release via air and water.
- 12. Population dose ccrinitments are
- 12. Both scrap and waste shipmnts should I
calculated for transportation of be r@m<4 when the plant expansion uraniun bearing materials, in-becoms operable.
cluitng scrap and waste.
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LIQUID WSTE DISPOSAL PPOPOSAL I
Uhe follcwing plan describes a proposed method for recovering uraniun contained in a CaF natrix in on-site waste lagoons. The concept includes renoval of ' sludge 2
frczn the lagocns, transfer to a mixing tank to solubilize the uraniun, " naturalize" the material by addition of depleted uranium and appropriately sample the material I
to verify that a natural uranien equivalency has been achieved.
i i-This material will then be used as feed to an operation which extracts uranium fran natural phosphate deposits. This operation will recover essentially all of the uranium and fluorides as elable products and dispose of the calcium as gypsutn (CaSo ).
4 Liquid waste treatment operations at Colurbia have generated approximately 2.5 matrix.
Table I million gallons of sludge containing uraniun in a CaF2 presents the physical and chenical properties of the CaF. This material will 2
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be transferred fran the lacoons to a tank where it will be heated and mixed with 2
l sulfuric acid and diatcxraticus earth to solubilize the uranium.
i Depleted uranium in an aqueous form will be addM to the slurry in sufficient i
quantities to " naturalize" it to a maximum 0.71% U-235 enrichm nt, adjusted dowruard to account for the sampl.ing error.
Once the naturalization step has been confirmd, the natural mixture will be transferred to storage tanks. Frcza these tanks, it will be transferred to tank trucks (ncminal 4000 gallon capacity) and shipped to Iakeland, Florida where the material will be introduced into the uraniun recovery operation at a rate of approximately 4000 gallons per day. At this point, the material will beccme an integral part of the uranius recovery unit process.
Key elenents of the operation inclixie:
assuring a hcmogenous slurry in the lagoons 1
i radiological controls and surveillance during lagoon and tank transfers to minir,ize contamination
sampling ard analyses of the naturalized slurry to ' assure uniform con-centration and thus una. form enrichment ccatrol of each shipnent batch to a madmum U-235 quantity 4
minimize separation of solids frcm liquids verify naturalization of the mixture selection of a transportation mode to safely transport tre material Environmentally, the proposal is very attractive when ccnpared with &a alter-native of dispasal of the sludge in an NFC licensed burial ground. E'irst, the bulk of the uraniun will be recovered for reuse in the light water fuel cycle.
l Second, the fluorides will be recovered as fluorosisilic acid. The only waste j
product will be gypsun which is a normal waste frcm the phosphate mining I
j operations.
f Approxunately 1,000 tank truck shipnants are required to reTove the existing a
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iiiventory.
If the program is successful, routine shipnents will be made to process the 550,000 gallons per year of sludge which is generated annmily.
l Based upon measured external radiation levels frcm the sludge, the pcpulation dose ccmnitmant is e.W to be minimal.
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A test program is planned to detenaine whether the uraniua in the sludge can be cbnverted to the soluble form and whether this material is chenically con-sistent with the phosphate plant's systen.
'Ihe follcwing licensing actions are requested:
1.
Westinghouse is currently licensed to perfonn waste processing, including acid treatment and dissolution, within existing facilities. We intend to 4
. perform the sludge preparation in spmin11y designed equipT.nt located near i
the waste lagoons. Def ails of this installation will be. supplied at a later date.
1
~emm
-mmn..-
2.
Approval is requested for tne downgrading of the eaa. sting sludge fran the present enrichmnt (approximately 2.71%) to natural. Minimum criteria for l
the downgrading is requested, including a definition of natural uranium and maximum acceptable sa:Tpling error. A detailed sampling plan will be supplied 2
at a later date.
=
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Approval is requested to remve this material fran NFC jurisdiction and i
transfer to the State of South Carolina jurisdiction. Subsequent approvals for material transfers, transport and recovery will then be pursued at the 1
4 appropriate state levels.
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PHYSICAL AND CHEMICAL PROPERPIES OF CaF2 6
1.
2.5 X 10 gallons by end of 1978 2.
3.05 X 106 gallons by end of 1979 3.
550,000 gallons per year ( 1500 gallons / day at 800 MIU/yr.)
4.
Properties High
' Im l
ph 12.5 10.1
% Solids 38.6 35.5 1-
% Water 61.4 64.5 I
U, ppn 286 152 U Enrich, %
2.71 Soluble F, ppn 33 27 i
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Major Impurities (Wt. %)
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l Al 0.7 0.3 f
Si 2.0 0.6 t
j Fe 0.3 0.3 l
Mg 1.0 1.0 i
i Ti 0.03 0.03 i
Ni 0.02 0.01 Mn 0.01 0.01 Cu 0.003 0.001 I
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REVISION RECORD
(
tevision Date of No.
Revision Paces Revised Revision Reason
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4 12-30-77 223, 224, 225 Revised criteria to agree with 0.3 fraction critical.
,-(
4 12-30-77 226 Data deleted.
4 12-30-77 230.1 Page added.
4 12-30-77 231 0.3 fraction critical.was 0.4.
(
4 12-30-77 234 Clarified requirements for indepen-dent review of data, i
4 12-30-77 254 Clarified retention of records of component approvals.
i 5
2-6-78 208 Revised to read "open face and labor-l atory-type hoods."
5 2-6-78 208.1, 209 Deleted airborne concentration option.
i 5.
2-6-78 228 Revised to qualify concrete reflection and calculation assumptions.
6 3-3-78 22 Expanded description of systems l
provided with emergency power.
6 3-3-78 24 Specified that waste processing is 4
conducted in the Contamination Con-trolled Area.
1 6
3-3-78 100 Specified that liquid waste evaluation is conducted in the Contamination Controlled Area.
6 3-3-78 210 Inserted word " automatically."
235 1 7 3-28-78 204 U increased to 50,000 kilograms
- 7 3-28-78 263.1 Revised to update bicassay frequency data.
l 7
3-28-78 263.2 Revised In-vivo action levels.
7 3-28-78 263.3 New table on urinalysis action levels.
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3-28-78 263.4, 263.5 Pages renumbered 8
3-5-79 18 Revised Figure 1.3.2.1 1
8
'3-5-79 20 Revised to include side entrance to 1
facility 8
3-5-79 21' Revised to include side entrance to l
facility Docket 70-1151Defe: 8-28-74 Revision No.
'8 Octo: ' 3-5 -7 9 '
Pooe 6
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REVISION RECORD Revision Date of No.
Revision Pages Revised Revision Reason 8
3-5-79 76 Revised to include solvent extraction to scrap treatment and to delete the last sentence of the third paragraph.
8 3-5-79 77 Revised Figure 1.9.0.1 to reflect plant expansion.
8 3-5-79 122-122b-Delete "1.9.4 Reserved" from middle
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of page. Add Figure 1.9.4.1 and Table
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1.9.4.1 8
3-5-79 151 Revised to include solvent extraction to scrap treatment and to delete the last sentence of the third paragraph.
8 3-5-79 153 Delete " Proprietary Information" at top of page.
8 3-5-79 158 Delete the words the material" on third and fourth lines and replace with "and purify the materials if needed."
8 3-5-79 160 Add the words "or UO2" to "U308" be-tween the blocks labeled " Thermal Processing" and Packaging and Storage."
8 3-5-79 164 Modify section 1.9.4.5 to read " Var-ious solid materials which are un-contaminated may be introduced into a thermal processing step directly to convert them into desired uranium oxide form."
8 3-5-79 194 Delete section 1.10 entitled "Off-Site Release Evaluation."
8 3-5-79 194a-k Add new section entitled " Auxiliary Incineration System."
8 3-5-79 1941-q Incorporate new section entitled
" Chemical-Manufacturing Development Laboratory."
8 3-5-79 194r-ae Incorporate new section entitled
" Purification of Contaminated Scrap Through Solvent, Extraction."
8 3-5-79 194af Reprint section 1.10 entitled "Off-Site Release Evaluations" previously printed on page 194.
8 3-5-79 206 Modify the first sentence in subpara-graph 2.2.1 to read "Four emergency stanby generators with rated capacity of 250 kw, 300 kw, and 2 at 500 kw will be maintained.
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Docket 70-ll510 ate: 8-24-74 Revision No. 8 Date:
3--5-79 page 7
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SfC4-1107 ELECTRIC SUB-STATION CALCIUM FLUORIDE LAGOON #3 (PP,0 POSED)
TANK FARM ALCIUM FLUORIDE CALCIUM FLUORIDE LAr-00N #2 LAG 00ii el MATER TANK (STOPJ.G STOPJ1 RU:10FF TO PUMP HOUSE d lis UPPER SUNSET LAKE l
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UF6 STORAGE PAD SANITARY LAGOON y
SHIPPIf!G CONTAINER STORAGE AREA 7
" BUTLER" BUILDING OIL HOUSE NORTH & SOUTH
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DITCH
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PROCESS WASTE t \\\\
e LAGOONS f
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FRAME HOUSE 7]
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EXPANSI0f!
EAST LAGOON COOLINGk I
.EQUIPliENT SHED WASTE TREATMENT BUILDIN SCRAP STORAGE
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AREA
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PLANT l
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BOILER HOUSE G
J CONTRACTOR CHAf;GE CAFE - "
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ROOM OFFICE O
ADVANCEB. WASTE
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TREATMENT BUILDING
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(PROPOSED) o$
VISITOR
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PARKING PARKIflG HF STCRAGE TAliK U*lH STORAGE TAf!KS DITCH 8
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150 300 450 600 HHHH -
i SCALE OF FEET COLUt1BIA SITE BUILDING LOCATIONS
,i FIGURE 1.3.2.1 1
Docket 70-1151 Date: 8-24-74 Revision No.1 Dofe:
2-6-79 Page
SNM-1107 1.3.4 (continued)
Access to the facility during normal working shifts is granted to:
I Ij (1)
Certain NFD employes and certain other Westinghouse employes assigned to the Columbia Site, through the front main entrance upon or second r
8 side entrance upon display of their identification badges to the security force on duty.
(2)
Other NFD employes to the NFD offices directly, upon display of their identification badges to the security force on duty at the front entrance or second side entrance of the facility.
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(3)
Other Westinghouse employes assigned to the Columbia Site through the front entrance to the facility.
Such visitors must sign a visitor's register and be issued a visitor's badge by the l
person on duty.
i (4)
Other visitors by registering at the appropriate l
NFD reception desk.
All visitors must be issued I
a visitor's badge and all non-Westinghouse visitors i
are escorted.
l (5)
Vehicles (as far as the shipping-receiving docks) j through electrically-operated gates controlled by i
the security force.
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Dockel 70-1151 Date: 8/28/74 Revision No. 8 Date 3-5-79 P, 20
SNM-1107 1
1.3.4 (continued)
During off-shifts, the entrances are locked and the facility is regularly patrolled at intervals not exceeding four hours by a member of the site security force.
All access during these shif ts is through the front entrance or second side entrance to the facility and is controlled by a member 8
of the site security force who is stationed there at all such times.
A security service report is initiated on a shift basis by the security guard and is kept on file by the Manager, Security and Services.
Keys for the locked doors controlling access to the Columbia facility are issued to the NFD Manufacturing /
i Columbia Plant Manager and to the Manager, Security and Services.
The site security force is issued keys for locked doors controlling access to the NFD Manufacturing building.
The Facilities Engineering Manager and the Maintenance Supervisors are issued keys to the factory l
area of the facility.
I Additional security measures as specified in Amendment SG-4, as amended, are implemented when mixed oxide fuel i
l is possessed on-site.
1.3.5 Utilities and Services 4
The Columbia Site is served by a single electrical supply x
line.
Four diesel-powered standby l8 generators are installed to meet the emergency electrical power requirements of the site in the event of a temporary 21 Docket 70-1151 Date: 8/28/74 Revision No. 8. DatE[
3-5-79' Pope l
c.
SNFt-1107 1.9 Processino Ocerations The processing operations to be performed on radioactive materials under this license may be divided into a number of distinct categories.
The first category is the ADU conversion operations
.which employ chemical means to convert UF t uranium oxide-6 powder.
The second category, then, is the fabrication operations which use essentially mechanical processes to produce fuel assemblies containing encapsulated UO2 pellets.
Another category is the analytical operations which use a variety of spectrographic and wet chemical operations on small samples of material to assure that material specifications are met.
Still another category is the treatment of scrap generated on site to permit it to be recycled into production operations or more closely controlled prior to discarding.
This treatment includes chemical dissolution and precipitation, solvent extraction and dry processes.
8 Westinghouse also has developed and is currently operating an alter nate process for converting UF to uranium oxide.
This is the Direct
- 6 Conversion Fluidized Bed (DCFB) process which may contribute to reduced chemical effluent levels.
Docket 70-1151 Defe: 8/28/74 Revision No. 8 D a t's-3-5-79 Poce 151
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SNM-1107 8
1.9.1.2 UF to UO C nversion 6
2 The chemical operations required to convert UF to uranium oxide are carried out in a con-6 tinuous processing line which is a closed
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system.
Engineered safeguards are
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applied to particular operations as required to prevent, or to detect and control, abnormal conditions.
Low enriched uranium in a closed system requires only routine radiation protection precautions.
The engineered safeguards specified above will effectively minimize the possibility of a significant material release.
Individual items of processing equipment are designed within the MPV for diameter, slab thickness, mass or
- volume.
For subcrits that are, or could credibly become moderated, the equipment spacing is established using surface density or solid angle criteria.
The system would remain suberitical for the maximum i
235U enrichment authorized for Model 30A or 30B cylinders.
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l.9.1.3 Storage of UO Powder 2
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The dry uranium oxide powder is stored in a closed system until it is released by Quality Control for use in the fabricating areas or for shipment in licensed packages to other facilities.
Low enriched uranium in a closed system or sealed packages requires only routine radiation protection Revision No. 8 Defe:
'Poge 153
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Docket 70-ll510CM 8/28/74
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SNM-1107 Scrap Recoverv Operations 1.9.4 1.9.4.1 General Scrap recovery process operations are characterized as batch operations involving a variety of input
/
and purify 8 The preliminary operations concentrate i
forms.
it to forms readily pro-the materials if needed and convert Not all materials require proces-cessed into U 0 powder.
3 sing through the entire sequence of operations.
f The basic processing sequence includes dissolution o i
ting solid forms, conversion to slurry form by precip ta ADU from the solution, dewatering the slurry form by wet mechanical separation, calcining the resulting sludge in furnaces, and packaging and storing the i
is sampled and j
resulting product.
The product analyzed before release to manufacturing.
i Various other inputs are fed into the basic sequence Liquids containing disolved at appropriate points.
1 uranium are introduced into.the solution hold tank Clean used to supply the precipitator columns.
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aqueous suspensions from mopping or off-gas
- filters, scrubbing operations and from laundering cloth poly bags, etc. are introduced prior to the wet Clean powder scrap, mechanical separation operation.
scrapped pellets and other quality solids may be subjected to a dry mechanical separation step and then sludge, to the calcinipg are introduced, along with the vet operation. Thermal processing capabilities also exist to 8
j convert this clean scrap material directly into usable l'02 158 j
powder.
Revision No.8 Datet 3-5-79 Poce Docket 70-1151 Date: 8/28/74
- kOCi";S CAT'ZOE'*
INPUT MAT;3 TAT, 1.
Equip. ment Clean-out AcM 7
g 2.
UN:1 CryuLulu
'Diccolutica To ADU
____w U:GI Colution Lines (Optional)
Chemical 3.
Conversion Area Clcan-out Precipitation B
Solution Cartridge Filters ADU Slurry e.
Washing UO 5.
Filter Precc Cloths Operations 4
C 6.
Polybags I
7.
Dust Collector Bags 9
V
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Wet 8.
Mop Water Mechanical D
9.
Scrubber Clean-out x
Separations ADU, UO2 U. 6 - 91ndee
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3.
Conversion Area Solid Clean-out ADU, UO,
Thercal
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2 U038
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11.
Pellet'Arca UO ' U3 8 2
12.
Floor Sueepin;;o Dry Mechanical F
13.
Absolute Filters Secarations U0 38
(
or UO 8
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.4.
Tank Vents Misc.
I C
Operations
- 15. Waste Solutions y
To Waste y
y Packaging &
3 Storage H
I, SCRAP ' *OVERY ORCCESS~ OPERATIONS ~ FLOW SHEET Figure 1.9 A.1 Docket 70-1151 Date: a /28 /74 Revision No. 8 pore:
3-5-70 Pone _ 161
SNM-1107 i
1.9.4.5 (continued)
Various solid materials which are uncontaminated may be introduced into a thermal processing step directly to convert them into the desired uranium oxide form.
1.9.4.6 Dry Mechanical Seoaration (Catecory F) 4 l
Non-homogeneous uranium oxides are crushed to a powder j
in a mechanical, granulator prior to the dissolution operation to promote the speed of the reaction.
Contaminated UO2 and U Og such as that 3
j found in floor sweepings is screened in a vibratory
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separator to remove trash and gross impurities.
In addition, accumulations of chemically uncontaminated g
UO or U 03 8 such as those occurring on absolute filters 2
t 1
in the Controlled Area are removed for further process-t ing by manual techniques such as shaking, scraping, I
etc.
Dry materials charging, dry mechanical separation, 4i pellet and powder granulation, and final packaging are 1
l performed in hoods, hopper dryboxes, 6r similar
!j enclosures equipped with individual blowers and HEPA l
j filters to minimize airborne contamination levels.
The filtered air from the enclosures is exhausted back into i
the Controlled Area.
Continuous air sampling heads are o
strategically located throughout the scrap recovery area to obtain representative samples of " breathing zone" air.
DOP tests and face velocity measurements 164 R*vi8 Ion No. 8 Date: 3-5-79 Pope Docket 70-1151 Dato: 8/28/74
i SNM-1107 1.9.6.8 (continued) indicated by the differential pressure monitoring device on the incinerator control panel.
It will also be indicated by the analyses of the stack samples.
If this occurs, immediate shut down of the incinerator will be initiated manually so that appropriate corrective action can be taken.
As enumerated above, sufficient means are provided for immediate detection and correction of problems that could occur in the incinerator system.
For this reason, adverse in-plant and offsite effects due to system failures are not considered likely.
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j 0-1151 8-24-74 Docket Dafe:
Revision No. 8 Date: 3-5-79 pog, 194 2
1
e SNM-1107 9
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1.9.7 Auxiliary Incineration Svstem 1.9.7.1 Purpose of Incineration Combustible waste may be treated by incineration to reduce the volume of waste disposed to licensed burial grounds and to permit the recovery of SNM j
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uhen practical.
1.9.7.2 Typical Materials to be Incinerated Typical materials are paper, plastic shoe covers, gloves, mops, plastic bags, tape, fiberboard contain-ers, burnable liquids, etc. from licensee's various laboratories and fuel fabrication processes.
1.9.7.3 Incinerator System Location The auxiliary incineration system is located in the Contamination Controlled Area of the Westinghouse Columbia Plant.
The specific location is shown in Figure 1.9.0.1.,
Revision 8 (page 77).
1.9.7.4 Incinerator System Description and Operation The incinerator is a controlled air, dual chambered, gas fired unit.
The two chambers, ignition (lower) will operate at approximately 1500 F and the (upper) combustion chamber will operate at approximately 2000*F.
In addition, the ignition chamber is equipped with a combustible liquid burner.
There is also a continuous Ash Removal System opening at the rear of the ignition chamber.
Exhaust gases leav.'.ng the upper combustion chamber will be transferred into a Quench Tower.
Docket 70-1151 Date:
8-24-74 Revision No. 8 Date:
3-5-79 Pope 194a
SNM-1107 l.9.7.4 Incinerator System Description & Oceration (cont. )
The exhaust gases are sprayed with conditioned water and are condensed and cooled.
This condensed liquid is adjusted for pH and reused in the gas scrubbing system.
/
Material movement through the auxiliary incineration system is described as follows:
1.9.7.4.1 Boxes of contaminated waste, with the amount of contamination recorded, will be delivered to the incinerator.
1.9.7.4.2 The boxes will be fed via a conveyor into the incinerator feed system.
1.9.7.4.3 The incinerator feed door will raise auto-matically as a hydraulac RAM pushes the box into the ignition chamber of the
~
incinerator.
~
1.9.7.4.4 The feed system can be adjusted from four (4) boxes to twelve (12) boxes per hour.
1.9.7.4.5 Live steam may be, injected into the chamber to assist control of the combustion pro-Cess.
1.9.7.4.6 The gaseous products of the combustion at approximately 2000 F will enter the Quench Tower where the temperature will be lowered to approximately 160*F.
1.9.7.4.7 The gaces will pass through a Venturi Scrubber section and into a hcl Acid Stripper (packed column).
IIbcket 70-1151 Defe: 8-24-74 Revision No. 8 Dofe:
3-5-79 Pope 194b
SNM-1107 1.9.7.4.8 The' gas flow then travels through a con-denser and any liquid removed is returned to the scrubber pump.
1.9.7.4.9 The dewatered exhaust gases are reheated by the duct heater before entering the
/
HEPA filter house.
All duct work between the condenser and the exhaust blower will be heated and insulated.
The top, bottom and sides of the HEPA filter house will also be heated and insulated.
1.9.7.4.10 The exhaust blower will be mounted in the second floor equipment room with a stack up through the roof.
An isokinetic probe will be installed a minimum of 5 duct diameters above the blower.
A small back-up blower will be installed in parallel with the primary blower.
The back-up blower will only operate when the primary blower fails.
The back-up blower is only to permit an orderly shut-down.
1.9.7.5 Radiological Safety Control The incinerator system is installed within the Contam-ination Controlled Area of the plant.
Only authorized personnel are allowed into this area.
Operating personnel are required to submit to the bioassay program for routine urinalyses.
Lung burden determin-ations (subparagraph 3.2.3) and the use of external radiation exposure monitoring devices (subparagraph Docket Date:
Revision No.
Defe:
Pope 194c
SNM-1107 1.9.7.5 Radiolocical Safety Control (cont. )
2.2.3) may also be required.
An isokinetic air sampler, described in subparagraphs 1.9.6.4 and 1.9.7.4.10 is installed downstream of the
/
HEPA filters in the incinerator filter house.
This
/
air sampler continuously collects samples representa-tive of the exhaust effluents discharged to the atmos-phere.
The samples are analyzed daily during operations.
If the exhaust effluent at the pcint of release
-12 reaches a level of 2 x 10 microcuries per milli-liter, an investigation will be made and the results
-12 evaluated.
If the investigation reveals that 4 x 10 microcuries per milliliter may be exceeded as an annual average ccncentration, the cause will we deter-mined and corrective action taken.
These release and action limits are consistent with those applied to all other exhaust effluents.
Effluent exnaust concentrati( as from the existing incinerator have averaged less t'aan 20% of MPC.
The exhaust system for the new incinerator is expected to represent an improvement over the existing incin-erator and thus lower exhaust concentraticns even further.
Air sampling requirements for the incinerator will be evaluated in accordance with subparagraph 2.2.6 and 3.2.2 o? this license.
Permanently mounted continuous air sampling stations will be established Docket 70-1151 Date:8-24-74 Revision No. 8 Date: 3-5-79 Page 194d
SNM-1107 1.9.7.5 Radiolocical Safety Control (cont. )
where the greatert concentration of airborne activity is expected under adverse circumstances and consis-tent with operator work locations during the initial j
loading operations, and during ash removal operations.
/
'/
The exhaust blower maintains the combustion chambers at a negative pressure with respect to atmospheric.
Under normal conditions, the combustion chamber will operate at a negative pressure of 0.1 ' H O or more.
2 The exhaust system will be maintained (filter change, scrubber maintenance, etc.) to assure that this mini-mum negative pressure drop of 0.1 " H O is maintained.
2 This should be sufficient to ensure adequate contain-ment since the combustion chambers are relatively air tight to assure that a proper combustion atmos-phere is retained.
Minimum instrumentation and con-trols are described in subparagraph 2.2.12 of this license, and apply to the auxiliary incinerator.
Incinerator ashes are continuously transferred from the combustion chamber directly to a ventilated enclosure.
This is performed within containment under negative pressure.
After transfer to an approved container, appropriate precautions will be exercised during removal of the containers to minimize airborne radioactivity.
Typically, all ash handling will be conducted within a ventilated enclosure designed to meet the requirements of subparagraph 2.2.5 of this license.
Docket 70-1151 Defe: 8-24-74 Revision No. 8 Dofe-3-5-79 Page 194e
SNM-1107 1.9.7.5 Radiolocical Sa'fety Contrcl (cont.)
At any time, liquids from the scrubber can be manually or automatically transferred to the final filtration tank where it is combined with conversion operation process wastes. These materials are then pumped through an on-line monitor system uhcre uranium content is determined and if within acceptable limits is pumped to the plant waste treatment facility.
Here they are refiltered and discharged to the environment in accordance with the limits specified in 10CFR20.106.
Due to the increased efficiency and improvements on this new system, total discharges of wastes are ex-pected to be decreased.
1.9.7.6 Nuclear Safety Control The area is monitored by a criticality alarm station which is part of the plant system and alarms both throughout manufacturing areas and at the plant Health Physics Laboratory.
The incinerator is operated on a nuclearly safe batch basic in accordance with the maximum permissible values of mass listed in Figure 1
2.3.2.1.
The enrichment used to determine the i
maximum permissible mass will be the highest value assigned to any of the wastes accumulated for incin-eration.
All associated equipment, such as the scrubber pump tank, heat exchanger, filter housing in the scrubber re, circulating system, and the vacuum cleaner t
to be used for ash removal is designed for processing wastes having a maximum enrichment of 4.15 w/o 235 U
Docket 70-1151 Date: 8-24-74 Revision No.8 Date: 3-5-79 Pope 194f i
SNM-1107 1.9.7.6 Nuclear Safety ' Control (cont. )
However, if it becomes necessary to process wastes 235 having enrichments between 4.15 w/o 0 and 5.0 w/o U,
this equipment, (if required), will be resized and respaced in accordance with the applicable criteria 7
/
given in Figures 2.3.2.1 through 2.3.2.4.
In addition, the batch size will be appropriately reduced.
Combustible wastes delivered to the incinerator area are contained in 30-55 gallon, metal or fiberboard drums, or fiberboard boxes presently used for baled waste shipments.
These containers of waste are scanned in a gamma counting facility.
After scanning, the containers are stored in an area adjacent to the incinerator, under established criteria for the Con-tamination Controlled Area.
The applicable storage criteria are those described in Paragraph 1.3.7.
Wastes accumulated without regard to origins are assumed to contain uranium having a maximum enrichment, normally 4.15%
U.
Proper storage of this material and storage area posting requirements are monitored routinely by the Plant Criticality Engineer.
Waste accumulated with known origins or enrichments are assigned the known enrichment.
They are stored correspondingly segregated in accordance with estab-lished operating procedures.
Stcrage areas for these wastet are reviewed and approved by the manager of the Radiation Protection Component.
Docket 70-1151 Defe: 8-24-74 Revision No. 8 Date: 3-5-79 Pope 194c
SNM-1107 1.9.7.6 Nuclear Safety Control (cont.)
All containers are marked with grams of U contained therein as determined by licensee's gamma counting facilities.
Each container is limited to 350 grams 35U reduced by 50% to reflect the measurement un-certainty associated with the gamma count.
The gamma counting facill. ties are calibrated at least monthly, and checked daily during counting operations, 235 using phantoms loaded with known quantities of g,
Size and geometry of the phantoms are identical to those of the waste containers.
An incinerator log is maintained by the operator, in-235 dicating the U content of waste charged to the system and ash removed from the system.
Until the limit specified below is reached, waste can be charged.
When the incinerator system is cleaned and the ash removed, the nuclearly safe containers of ash are gamma counted, the SNM content recorded in the log and a comparison made with the amount charged.
The difference will be considered MUF and recorded in the l
l log.
Ash may also be analyzed chemically to verify i
235 i
the U content.
A particular batch portion is not charged when the I
j sum of such an additional charge plus the recorded e
quantity of MUF plus the amount already charged ex-j l
ceeds a nuclear safe batch of SNM reduced by 50% to reflect the measurement uncertainty.
When this limit
)
Docket 70-1151 Date: 8-24-74 Revision No. 8 Dole: 3-5-79 Poge 194h 1
SNM-1107 1.9.7.6 Nuclear Safety Control (cont.)
reached, incinerator operations are suspended, ash removed and gamma counted, and a survey of the system made to measure any biases which may exist between the feed and ash counting steps.
The recirculating scrubber solution can be sampled anu analyzed by alpha proportional counting techniques.
When loaded, HEPA filters are gamma counted in a similar manner as feed material.
In addition, certain sections of the system may be surveyed visually, or with direct reading portable instruments.
Alternatively, the MUF may be adjusted to zero after each burn.
In this case, the ash would be removed and the entire incin-erator system would be thoroughly cleaned, visually inspected, and surveyed for residual contamination prior to release by the Radiation Protection Component.
After operation of the incinerator, ash is removed with a vacuum cleaner or other means and loaded into containers such as polypaks, fiberpaks or metal pails.
Both the vacuum cleaner and the ash containers are either limited to a nuclearly safe volume or a nuclearly safe diameter.
The applicable maximum permissible values specified in Figures 2.3.2.2 and 2.3.2.3 respectively for 4.15 w/o homogeneous oxides will be used.
The ash is gamma counted and a compar-ison of ash count with charge count is made and a MUF determined.
Whenever the total MUF (initial plus adjustments) approaches the safe mass limit, Docket 70-1151 Defe:
8-24-74 Revision No. 8 Dofe:
3-5-79 Page,na;
SNM-1107 1.9.7.6 Nuclear Safety Control (cont. )
the system is thoroughly cleaned and, if necessary, new components are installed and the system is resur-veyed to establish a new system (initial holdup) MUF.
Each container of ash is limited to a maximum of 350 35 grams U re.duced by 50% to reflect the measurement uncertainty.
Containers of ash are stored in a desig-l nated section of the Contamination Controlled Area near the incinerator in accordance with the applicable nuclear criticality control criteria established in subparagraph 2.3.2.2.
Ash may be processed for recovery of the SNM or disposed of to a licensed burial l
facility.
l The wet scrubber system sump tank (water reservoir),
filters and heat exchanger are each in the form of cylinders with an effective inside diameter of 10.2" or less.
The heat exchanger is considered a flow through device and the sump tank and filters are spaced in accordance with surface density criteria.
l 10.2" is the maximum permissible cylinder diameter for 4.15 w/o material specified in Figure 2.3.2.3 for homogeneous material.
1.9.7.7 Safety Mechanisms Safety controls exist in several areas of the system 4
to insure safe operation of the system as well as t
i control operational upsets and/or malfunctions that could occur.
These are listed as follows:
Docket 70-ll51Date:
8-24-74 Revision No. ' 8 Date:
3-5-79 Page 194j
4
?
SNM-1107 1.9.7.7 Safetv Mechanisms (cont.)
1.9.7.7.1 High temperature (approximate 250*F) detected in _he Quench. Tower Sump. Alarm Indication.
Appropriate actions are taken to correct the situation.
1.9.7.7.2 Low indications in the scrubber system weir liquid flow.
Alarm Indication.
Appro-priate actions are taken to correct the situation.
1.9.7.7.3 Low indication on system air flow.
Alarm Indication.
Appropriate actions are taken to correct the situation.
1.9.7.7.4 AP HEPA filter high.
Alarm Indication.
Appropriate actions are taken to correct the situation.
As enumerated above, sufficient means are provided for immediate detection and correction of problems that could occur in the incinerator system.
For this rea-son, adverse in-plant and off-site effects due to system failures are not considered likely.
1.9.7.8 Improvements in the operation, logic and functional control of this system may be made as indicated by new regulations and/or operating experience.
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Docket 70~ 115 bate:
B-24-74 Revision No. 8 Date:
3-5-79 Page
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s 1.98 Chemical-Manufacturina Develorment Laboratory
- 1. 9. 8.1 Purcose of Chemical-Manuf acturine Develcoment Laboratory The purpose of the Chemical-Manufacturing Development Laboratory is to provide a separate area where development of processes and equipment can be acccmplished with minimum impact 'on normal production operations.
This i
area will be used for development, prototype development and equipment checkout prior to installation in the production operations.
i.
i l
1.9.8.2 Tvpical Operations That May occur In the Develocment Laboratorv The three typical areas of operation that may occur i
are listed below:
1.9.8.2.1 Chemical Development such as Waste Treatment Studies, Uranium Chemical Processing, Uranium Recovery, etc.
1.9.g.2.2 Ceramic Development such as Powder j
Preparation and Characterization, ij Pellitization Studies, Sintering Studies, i
etc.
l i
1.9.g.2.3 Mechanical Development such as Rod I
{
Loading Devices, Rod Plugging and 4
I Welding Development, etc.
I t
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- 1. 9.5. 3 Specific Examoles of Develocment Ocerations 1
The following is an example of a develooment croiect j
that was conducted at the Columbia fac.ility.
Docket 70-1151 Octs: 8-24-74 Revision No. 8 Date: 3-5-79 Page '19 41' 'd
SNM-1107 1.9.8.3 Specific Examples of Development Operations (cont.)
The development was conducted in one of the produc-tion lines during routine operations.
/
Docket 70-1151Date: 8-24-74 Revision No. 8 Date:
3-5-79 Page 19s,.,
SNM-1107 1.9.E.3.3 Pilot Furnace Testine Sintering furnace testing is currently performed on production line furnaces when production schedules permit.
The types of
/
testing conducted in the past included bind-
/
er studies, densification testing, etc.
Many of these programs are delayed because of production schedules.
Again, a development laboratory would permit furnace testing independent of production and under controlled conditions.
A develop-ment laboratory pilot sintering furnace would use enriched uranium under the same nuclear safety controls as production line furnaces, e.g. slab thickness and other controls to assure that loaded boats are not stacked beyond the allowable slab thickness.
Radiological safety will be assured by con-trol of hydrogen and appropriate exhaust ventilation through HEPA filters.
No unu-sual problems are anticipated.
1.9.@.4 Chemical-Manufacturine Development Laboratorv Locatic The Chemical-Manufacturing Development Laboratory is located in the Contamination Controlled Area of the Westinghouse Columbia Plant.
The specific location f
is shown in Figure 1.9.0.1, Revision 8 (page 77).
Docket 70-1151 Defe: 8-24-74 Revi:!an No. 8 Date:
3-5-79 Poce 194B _
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- 1. 9.8. 5 Chemical-Manufacturing Development Laboratory Descrip-tion The laboratory will occupy approximately 4,400 square feet total area of floor space.
It will be located adjacent to the Production Manufacturing Area and will be isolated by a wall or fence.
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Services such as ventilation, water chemicals, etc.
will be provided as needed.
Equipment and operations used for development will be evaluated by the Radiation Protection Component for compliance with existing license and regulatory requirements prior to operation.
The laboratory will be monitored by the area criti-cality alarm system.
The personnel bioassay program described in Section 3.2.3. will apply to persons working in this area.
1.9.8.6 Chemical-Manufacturing Development Laboratory Controls The scope of projects anticipated for this area is necessarily very broad.
This will require the follow-ing administrative controls to assure that radiological I
and nuclear safety concerns are addressed and appro-i priate controls are implemented.
l f
In this regard, lines of organizational authority for the laboratory follow the philosophy outlined in sub-paragraph 3.1 of the license, i.e.
line management is responsible for all aspects of the operations, in-cluding safety.
Docket 70-1151 Date: 8-24-74 Revision No. 8 Date: 3-5-79 Poge 1940
~~ i_
- 1. 9. 8. 6 Chemical-Manufacturina Dev. Lab. Controls (cont. )
The radiation protection function's responsibility is to review all development work from a radiological and nuclear safety standpoint.
All work involving radioactive materials require either an effective
/
detailed procedure or a job safety analysis.
Proce-dures are reviewed by the radiation protection function for radiological and nuclear safety.
Whether the work is performed under approved procedures or job safety analyses, the responsible line manager is required to submit sufficient information to permit the proper review.
The radiation protection function then issues the nuclear safety posting criterir. and inspects installed equipment as appropriate.
Air sampling requirements for the laboratory will be evaluated in accordance with subparagraphs 2.2.6 and 3.2.2 of this license.
Permanently mounted continuous air sampling stations will be established where the greatest concentration of dirborne activity is expected under adverse circumstancer and consistent with operator work locations.
swever, engineering controls will be used where possible to control airborne radio-activity at its source.
Ventilated enclosures are used as appropriate to con-trol airborne concentrations.
Examples include a walk-in ventilated hood for testing large pieces of equipment, small fume hoods, dust collectors, etc.
Docket 70-1151 Defe:
8-24-74 Revision No. 8 Dofe:
3-5-79 Pope 1949
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- - ~
.g 1.9.g.6 Chemical-Manufacturina Dev. Lab. Controls (cont. )
All enclosures will be designed and operated to meet the ventilation specifications in subparagraph 2.2.5 of this license; ventilation control procedures will conform to subparagraph 3.2.2 of this license.
Docket 70-1151 Dcfe:
8-24-74 Revision No. 8 Date:
3-5-79 Page 1947
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ir.
- 1. 9. 9 PURIFICATION OF CONTAMINATED SCRAP THROUGH SOLVENT EXTRACTION 1.9.9.1 Purpose of Solvent Extraction Contaminated uranium scrap may be purified through I
in-plant solvent extraction and thus virtually elim-inate the shipment of uranium-bearing scrap material
/ over the pub 1'ic highway system.
/
1.9.G.2 Typical Materials to be Purified Through Solvent Extraction Typical materials are uranyl nitrate solutions pre-pared utilizing the following scrap components: re-
__[~~~ quiring purification.
These include grease and oil
~
~ contaminated UO,,,
incinerator ash, oxidation furnace product, etc.
These scrap materials result from licensee's fuel fabrication process operations.
1.9.-93 Solvent Extraction System Location The solv 2nt extraction system is located in the Contam-i inated Controlled Area of the Westinghouse Columbia Plant.
The specific location is shown in Fiaura 1.9.0.1 Revision 8 (page 77),
1.9 44 Process Description and Eauipment l
1.9.9.4.1 Solvent Extraction System
'X 194R-Docket 70-115]Date: 8-24-74 Revision No. 8 Dofe: 3-5-79 Page
~
= -
SNM-1107 1.9 9 4.1 Solvent Extraction System (cont. )
The solvent extraction system consists of scrap preparatory equipment, dissolver, mechanically pulsed extraction and strip-pi, columns, product evaporator, a nitric
/
acid recovery system, associated liquid i'
storage vessels and related instruments and controls.
Scrap uranium is charged to the dissolver in preweighed critically-safe batches where nitric acid and water dissolve the uranium and forms a uranyl nitrate solu-tion.
The dissolver product is filtered, adjusted to a maximum of 5 gm U/ liter in nitric acid and pumped to the feed storage 235 tanks.
The concentration of U
is less 5 gm/ liter in all other vessels.
The dis-solver tanks, adjustment tanks, and feed storage tanks meet the MPV for diameter cri-teria.
All liquid transfer lines are less than 2 inches in diameter and do not require criticality consideration.
The process flow diagram is shown in Figure 1.9.sr. 4.1.
1.9 94.2 Preparatorv Ecuipment Prior to dissolution, uranium bearing scrap is treated to remove impurities such as vola-tile materials.
These operations will include thermal processing, solids removal anc^
Docket 70-1151 Date: 8-24-74 Revision No. 8 Date: 3-5-79 Page _19,4s
_ SOLVENT EXTRACTION SYSTEM FLOW DIAGRAM Figure 1. 9.9b 4.1
~n, I
Uranium Make-up l
{HNO3 Supply HO i
Scrap D.I.
2 L
')
v v
y i
I Dilute
)
- . Nitric Acid <
3 Head Tank Dissolve Head Tank
{ HNO n
mm V
Dilute i
s mm b
V v
v
,g Extract I i ~, HNO Con-
'O
~
SupplyTadk l
centr h
Acid Solvent-y g
MTH y
y Y
s 4
ft 3
Strip g
Utm Weak
)
}l o
Acid Degraded Solvent y
V Waste I
Acid Recoven Concentrate Very Weak Acid
-~
Hi Er.richment Wasce From HNO3 'Japori::er mlH Blend Supply se v
i Blend y
I i
Lo Enrichment UNH Blend Supply l
l Storagc UNH j
sr ADU Line j
s 4 Docket 70-1151 Octe: 8-24-74 Revision No. 8 Date 3-5-79 Page 19,4_t.-
SNM-1107 F.T S L ',
l.9.94.2 Preparatory Eauioment (cont. )
concentration of the scrap materials.
Thermal processing and concentration will be conducted in oxidation furnaces on a batch basis.
Solids removal will be incor -
porated into the dissolver system which is discussed below.
1.99 4.3 Dissolution Dissolution is performed in a series of nuclearly safe diameter tanks designed to sequentially dissolve the scrap in a nitric acid solution and settle out the non-dissolved solids.
The dissolved scrap is then trans-ferred to nuclearly safe diameter storage tanks where it is diluted prior to trans-fer to the extraction column.
1.99 4.4 Extraction Uranium is extracted from the acidified feed solution in a four inch safe diameter column containing spaced sieve plates and j
pulsed by a piston pump.
The extracting i
-~
tributyl solvent is a phosphate (TBP) mixture.
~~[
The solvent mixture is metered into the top of the column.
The uranium-bearing aqueous feed is charged in slightly below the column mid-point and flows toward the top of the column counter-current to the solvent mixture flow.
Docket 70-1151Date: 8-24-74 Revision No. 8 Dofe: 3_S_79 Page 194u
3-SNM-1107
~.
1.9.9.4.4 Extraction (cont. )
A nitric acid scrub solution flows into the bottom of the column and toward the top of the column combining with the aqueous feed stream in the upper region of the column.
Superimposed on these stream flows is an oscillation provided by the pulsing pump which forces the liquids back and forth through the sieve plate holes creating a dispersion of solvent droplets in the aqueous continuum.
The uranium transfers from the aqueous phase into the solvent phase aided by the " salting out" action of the excess nitric acid present.
The extracted aqueous phase collects in a " quiet zone" provided at the i
top of the column, overflows into a ten inch nuclearly safe diameter holding tank and is directed through a solvent trap to a nitric acid recovery unit.
A radiation detector monitors the column overflow and automatically shunts the stream into a ten inch nuclearly safe diameter auxiliary tank for recycle to 1
the feed tanks if a high uranium content is indicated
(>
1 gram U-235 per liter).
The 7
uranium lade.2 solvent phase collects in a quiet transition zone provided at the bottom i
of the column and then flows to the top of the stripping column.
Docket 70-1151 Date: 8-24-74 Revision No. 8 Dofe: 3-5-79 Poge '1942n
~ - -.
1 SNM-1107 a~
b
- 1. 9.9. 4. 5 Uranium Stripping The uranium laden solvent from the extraction column flows into the top of a six inch safe diameter column containing spaced sieve plates and pulsed by a piston pump.
The solvent phase flows toward the bottom of the stripping column counter-current to a stream of weakly
'. stripping acidic nitric acid.
/
water that is metered into the bottom of the column.
Superimposed on the flows of these two streams is an oscillation provided by the pulsing pump which forces the liquids' back and forth through the sieve plate holes creating a dispersion of solvent droplets in the aqueous phase.
The uranium transfers from the solvent phase into the stripping water.
The uranium laden stripping water i
l collects in the transition zone provided at the top of the co'lumn and overflows through p
a solvent trap into a ten-inch nuclearly safe diameter holding tank.
This uranium solution contains (35-100) gms.'
U/ liter l
The stripped solvent collects in a transition zone provided at the bottom of the stripping I
' column and flows into a ten-inch nuclearly i
safe diameter cylindrical holding tank.
f From there it recycles to the extraction 1
column again to extract uranium from a new increment of feed.
Dockel 70-1151 Date:
8-24-74 Revision No. 8 Date: 3-5-79 Pope 19 4w-
SNM-1107 i
1.9 9 4.6 Uranyl Nitrate Concentration Dilute aqueous phase product from the stripping column is transferred into a safe diameter, cylindrical, flash evaporater in order to remove excess water from the uranium solution.
The evaporated water is condensed and recycled to a storage tank for reuse in the extraction process.
The uranyl nitrate product, concentrated to a maximum of 5 gm U235/ liter, is pumped from the evaporator to product holding tanks (10 inch nuclearly safe diameter cylinders)
~
where it is sampled for isotopic content, acidity (as free HNO ) and uranium content.
3 Adjustments to acid content and isotopic content (when needed) of the product are continuously made as the product moves through a ten-inch safe diameter cylindrical mixing tank.
The material is then sent to J
f product storage tanks for use in the uranium 5
conversion lines.
i
- 1. 9. 9. 4. 7 Nitric Acid Recovery I
The aqueous effluent from the extraction column is transferred into a nuclearly safe diameter cylindrical evaporator which vapor-izes almost all of the liquid.
A small 3
amount of liquid is necessary to keep the Docket 70-1151 Date: 8-24-74 Revision No.
8 Date: 3-5-79 Pope 194x
WESTI,NGHOUSE PROPRI,ETARY CLASS II SNM-1107
- 1. 9. 9. 4. 7 Nitric Acid Recovery (cont. )
metal salts carried by the effluent in a slurry solution.
The vapor stream is piped into a safe diameter distillation column which separates the nitric acid as approxi '
/
/
mately 68% acid and this is returned to a storage tank for reuse in the uranium scrap dissolver.
The distillation overhead, which is essentially water, is recycled to a storage tank for reuse in the extraction process.
1.9 94.8 Process Waste Treatment Tank vents are connected to the plant vessel ventilation system which controls atmospheric pullutants to the lowest practical level.
Non-condensable off-gas from the uranyl nitrate evaporator and nitric acid still are removed through the plant vessel ventila-tion system.
The vessel ventilation system contains a direct contact, recirculating, j
venturi scrubber with a liquid disengagement t
section followed by a demister and finally, i
HEPA filters prior to discharge of gases to i
the atmosphere.
Filtered gases discharged to the atmosphere are continuously sampled and are analyzed for airborne particulate radioactivity on a daily basis.
Scrubber Docket 70-1151 Date: 8-24-74 Revision No. 8 Date: 3-5-79 Poc' 19 1 _
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SNM-1107 1.9 9 4.8 Process Waste Treatment (cont. )
water is transferred to the final filtration tanks where it is combined with conversion operations process wastes.
These materials are then pumped through an on-line monitor system where uranium content is determined and if within acceptable limits is pumped to the plant waste treatment facility.
No routine liquid discharges from the solvent extraction system are expected to be processed through the plant waste treatment facility.
A periodic solvent renewal of approximately 30 gallons for each 300 hcurs of column operation is expected to be needed.
The spent solvent mixture is drawn from the holding tank into a drum and replenished witn fresh solven,t mixture.
I The preferred method of disposal of the j
degraded solvent will be to wash the spent solvent solution with a sodium carbonate solution to remove the degradation products and then reuse the solvent in the process.
l Two additional possibilities for disposal of the spent solvent exist:
(1) further salvage uranium from the spent solvent such that the material may be disposed of as Docket 70-1151 Defe:
8-24-74 Revision No. 8 Date: 3-5-79 Page 1942
-~~
SNM-1107 O'*
L.
1.9..fA4.8 Process Waste Treatment (cont. )
non-radioactive chemical waste, (2) absorb the liquid in vermiculite and ship to a licensed burial facility.
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The metal salt slurry residue from the nitric acid recovery vaporizer also contains traces of uranium.
This slurry will be treated with a sodium carbonate solution to further remove uranium from this material.
The remaining residue will be disposed of through burial.
The carbonate solution containing the salvaged uranium will be transferred to scrap recovery operations in order to recover the uranium for use in production operations.
All other streams from the solvent extraction process are internally recycled as described in previous paragraphs.
- 1. 9. 9. 5 Radiolocical Safety Control The solvent extraction system is installed within the Contamination Controlled Area of the plant.
Only authorized personnel are allowed into this area.
Operating personnel are required to submit to the bioassay program for routine urinalyses.
Lung burden determinations and the use of the external radiation exposure monitoring devices r
Docket 70-1151 Defe: 8-24-74 Revision No. 8 Date: 3-5-79 Poge M 4aa
~
,2-1.9.9.5 Radiological Safety Control (cont. )
may also be required.
Permanently mounted, continuous air samplers are located at points within the
/
system where greatest concentration of air-
/
borne activity is expected under adverse circumstances and the points where operators work during routine operation of the system.
All tanks, columns, enclosures and furnaces will be ventilated into the existing plant exhaust system.
All tanks and columns will be vented into the present scrap recovery scrubber system as described in subparagraphs 1.9.4.2, 1.9.4.5 and 1.9.4.7 (non-proprietary) of the license.
Treatment of these off gases include scrubbing,and HEPA filtration.
Ventilated enclosures used for containment of dry scrap will be directed through the existing plant exhaust system which includes HEPA filtration.
An average of 100 lfpm will be maintained across all openings.
Ventilation control procedures and administra-tive controls are described in subparagraphs 2.2.5 and 3.2.2 of the license.
Docket 70-1151 Date: 8-24-74 Revision No. 8 Dofe:
3-5-79 Pope 194ah
- 1. 9. 9.
5 Radiological Safety Control (cont. )
With the effluent treatment described above, no significant increase in exhaust concentration is expected.
- 1. 9. 9'. 6 Nuclear Safety Control
/
/
The nuclear safety of the system is assured on the basis of safe mass, concentratic..,
geometry, volume and by solid angle criteria.
The double contingency criteria is met by solid angle criteria and one or more of the MPV's for mass, concentration, geometry or volume.
The dissolver is identical with the ones now used in Dissolution Operations Section 1.9.4.2 (non-proprietary) of the license and are operated on a safe mass basis.
All solutions down stream of the dissolver are 5 gm U235/
liter so that neu. tron interaction is of limited concern.
All process vessels and equipment meet the MPV's for diameter criteria.
Solid angle analyses were performed using solid angle criteria as given in TID-7016, Revision 1, and the supplemental reflector i
conditions as specified in Subparagraph 2.3.2.2 of the license.
Figure 1.9.9.4 1 I
shows the layout and center element and table s
- 1. 9.,9.4.1 shows the center-to-edge separation Docket 70-11'51 Dde: 8-24-74 Revision No.
8 Dofe: 3-5-79 Pm19'4ac-
'N 1.9.9.6 Nuclear Safety Control (cont. )
distances between the center element and other process vessels and equipment.
Al-though the geometric centers of all the process vessels and equiptaent are not in
/
th'e same plane, all separation distances i
were conservately assumed to be at right angles with respect to the lines joining their geometric centers with that of the center element.
All subcrits were reduced to cylindrical geometries on an equivalent cross-sectional area basis using the formula w= (sins), from TID-7016, Revision 1.
No distinction was made between horizontally and vertically oriented subcrits.
That is, the orientation of all subcrits was assumeu to be vertical.
Solid angle calculations were made assuming three (3) different vessels V-1081, T-1087 C and T-1087 D to be the central subcrit.
The K f r each of the central subcrit eff vessels was determined to be 0.580 from the curve shown in Figure 2.3.2.13 of the license.
The solid angle for T-1087 C was the largest for the central suberits and was determined to be 2.5748 steradians which is 80.5% of the maximum allowable value.
The layout, 194ad Docket 70-1151 Date: 8-24-74 Revision No. 8 Dofe: 3-5-79 Page
i SNM-1107
.,e
't
- 1. 9. 9,. 6 Nuclear Safety Control (cont. )
therefore, satisfies the solid angle criteria.
The area is monitored by two criticality
~
alarm stations of the plant gamma alarm system with dual scintillation detectors and three siren type audible alarm horns.
l Tanks which exceed th: MPV for cylinder diameter criteria such as the HNO head 3
j tanks are vented to preclude an inadvertent vacuum transfer of uranium bearing solution I
from the process vessels to the tanks.
All process vessels are vented into a safe
~ ' '
geometry scrubber.
t I
l 1
3 l
i
\\
Dockel 70-1151 Date: 8-24-74 Revision No. 8 Date: 3-5-79 Pope 194aq
SNM-1107 1.10 Off-site Release Evaluations Westinghouse has postulated a variety of accidents in the handling of large quantities of source and low enriched special nuclear materials and has found the consequences of any of them to be well within established guidelines.
The most significant type of accident would be a release of an appreciable quantity of radioactive material off site.
A quantity of uranium m'ay be established, which, if released, would produce a downwind concentration at ground level equal to the maximum allowable concentration specified in 10 CFR 20.106 (a).
Docket 70-1151,,,;
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SNM-1107 2.2.1 (continued) 8 Four emergency standby generators with rated capacities of 250 kw, 300 kw, and two at 500 kw will be maintained. The standby generators will be activated automatically in the event of a facility power outage.
2.2.2 Emergency Equipment Equipment required to cope with a radiation emergency will be kept at designated locations and will be sufficient to provide emergency personnel adequate radiation protection to meet the requirements of 10 CFR 20 during corrective activities.
Film and/or TLD badges and pocket dosimeters capable of detecting and measuring gamma and x-radiation will be available to emergency personnel.
Portable instrumentation, which is available at various locations on the site for the evaluation of beta-gamma radiation, will have capabilities over the range of 0.1 mR/hr - 300 R/hr.
Personnel protective equipment, such as respirators, self-contained breathing apparatus and protective clothing; and other required equipment, such as signs, rope or tape for marking exclusion areas, smear papers, blank data forms, and floor plans of buildings including equipment layout will also be maintained.
2.2.3 Perconnel Monitorina Devices Film badges or thermoluminescent dosimeters (TLD's) provided by a commercial supplier, capable of detecting and measuring gamma and x-radiation will be used.
In addition, neutron detection capability will be-available when specified by the radiation protection function.
Revision No. 8 pop 3-5-79 Pope 206 Docket 70-1151 Defe: 8/28/74