ML19294B028
| ML19294B028 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 08/10/1979 |
| From: | Reitler E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 13809, NUDOCS 8002270026 | |
| Download: ML19294B028 (3) | |
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U. S. Nuclear Pegulatory Camission Office of Nuclear Material Safety & Safeguards gc Division of Fuel Cycle & Material Safety Washington, D. C.
20555 Attention: Mr. L. C. Ecuse, Chief Fuel Processing & Fabrication Branch Columbia site visit by Messrs. R.' L. Stevenson and A. L. Soong Paference:
July 17-18, 1979 concerning Smi-1107 application Gentlenen:
Trancmittal of naaiHnnal infor: ration to supmrt the acclication
Subject:
for amerernent to expand far i1ity During the above referenced visit, additional infonration was requested re-22, 1979.
garding tm questions which were contained in an NT letter datied May The Westinghouse response to Question 8 stated that TBP degradation products are routinely removed fram the solvent extraction systen prior to intrcduction into uranyl nitrate concentration and nitric acid recovery vessels. This is acecrrplished using a diluent washing technique which renoves TSP products from both the aqueous product and aqueous raffinate streams.
Westinghouse and NRC representatives discussed the proposed " naturalization" of liquid waste generated at the Colt:rbia site. Several questions wre raised concernin3 the process, related to homogeneity and sampling of this material.
The fe'bwing is a brief outline of the process Wich address these concezns.
Li. quid wastes frcra the conversion of UF6 to m2 are presently treated and trans-ferred to a chemical waste treatment facility which retaves anrcnia and fluorides frcm the stream. The uranit:n centent of the liquid waste as it leaves the manu-facturing plant is typically less than 30 ppn.
An advanced uranit:n rer:cval process is planned which will further reduce uranit:n concentrations in the stream to less than 1 pga. An aquecus strea:n of depleted
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Mr. L C, pouse Page 2 August 10, 1979 uraniu:n will then be added to this effluent in proportions necessary to re-duce the enrichment to 0.72% in U-235 or less, i.e. to "naturalire" the stream.
Uranitra contained in the liquid stream leaving the manufacturing plant is primarily in a soluble state (the insoluble mterials are romved by filtra-tion prior to discharge from the plant). Additional uranium is then reroved by the advanced process. The liquid effluent frcrn this process and the de-pleted uranium addal is in soluble form.
The depleted uranium will be in a form such as uranyl nitrate or uranyl sulfate solution containing approxinntely 0.5 gm/l total U at an ndaition rate of approximately 20 ml per gallon. At this rate, the total uranica concentration in the combined stream amounts to less than 10 p5xn which is not sufficient to cause ADU precipitation since the soluble uranium content in the ADU waste stream ramally amounts to 15-30 p5m.
The present Chanical Waste TreatInent Facility removes fluorides by precipita-follcwing the naaition of Ca (CH)2 to the liquid waste stream.
tion as CaF2 This material is then transferred to a Still where amtonia is recovered for reuse in the manufacturing plant. The Still battams are pumped to a settling is collected.
lagoon where the CaF2 Depleted uranium will be intr *mi prior to the Ca(OH)2 nadition step and will be thoroughly mixed with the liquid waste; the Ca(OH)2 will then co-precipitate all uraniun. The precipitate should also be hcxnogenous with resoect to enrich-Westinghcuse plans to corduct testing to demonstrate that the physical ment.
mixing is effective and that the precipitated uraniu:n is lxnogenous ard less than 0.72% in U-235.
A ccrgrehensive 9 - ling plan will be developed to assure that the above criteria are met. Representative sm'Tpling of the CaF2 precipitate will be performed either en a batch basis from process vessels or proscrtionately sampled frcm process trans-fer lines. Analyses of uraniun content and U-235 enrichnent will be performed.
The enrichment goal will be chosen so that systen fluctuations will not result in exceeding 0.72% U-235.
The required sampling frequency will depend upon (1) the desired level of confi-dence that the resulting enrichment will be below 0.72% U-235, (2) the chosen value for the " control enrichment level", and (3) the errors associated with analytical techniques. At a given level of confidence and analytical error, as the "centrol enrichrent level" approaches 0.72% U-235, a larger ntrrher of sanules will be required to produce that level of confidence that the nuxistra enrichment is not exceeded.
For this process, a 95% confidence level will be used. We will then select a "centrol enrichrent level" below 0.72% U-235 and a sampling fregaency to assure that the natural enrichrent level is not exceeded at the 954 ccafidence level.
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Mr. L. C. Rouse Page 3 August 10, 1979 The specific ntster of samples and the " control enrichnent level" will be established in the future when the advanced waste treatnent process design is empleted.
The expected analytic e.rror for enrichment detennination is eW. to be less than 53, including solution concentration regaired for rass spectro--
retry isotopic detenninations.
Once the rater.ial has been transferred to lagoons, periodic sampling will be perkrmac1 to verify unifonn uraniu:n distribution throughout the CaF2 matrix and that the enrichr.ent is less than 72% in U-235.
We trust that this _ nFn= tion is sufficient to ccuplete your revie'a of the i
above referenced application. If you need adaitional infouration, please telephone me at (803) 776-2610 Ext. 395.
Sincerely, 9ESTINGESE ELECI'RIC CORP.
E. K. Reitler, Fellcra Engineer Radiological & Envuomental Engineering EKR:le
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