Text

Second Partial Response to FOIA Request for 26 Documents Re GE Facility in Wilmington,Nc on Complaint Filed W/Dept of Labor.App a & Portions of App B Documents Available in Pdr. Portions of App B Withheld (Ref FOIA Exemption 4)
	ML20138G235
Person / Time
Site:	07001113
Issue date:	12/05/1985
From:	Grimsley D; NRC OFFICE OF ADMINISTRATION (ADM)
To:	Ratner M; RATNER, M.G.
References
	FOIA-85-158 NUDOCS 8512160270
	Download: ML20138G235 (5)
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~ ' UNITED STATES U ;

o NUCLEAR REGULATORY COMMISSION I

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Mozart G. Ratner, Esquire 1900 M Street, NW, Suite 610 IN RESPONSE REFER Washington, DC 20036 TO F01A-85-158-A

Dear Mr. Ratner:

This is the second partial response to your letter dated February 25, 1985, in which you requested copies of 26 specified documents related to General Electric Company's facility in Wilmington, North Carolir.a. with regard to a complaint you filed with the Department of Labor on behalf of your client.

This also addresses your letter dated June 27, 1985, regarding this matter.

The records identified on the, enclosed Appendix A and the released portions of the records identified on the enclosed Appendix B are being placed in the Nuclear Regulatory Commission's (NRC) Public Document Room (PDR). You may obtai.n access to these records by requesting PDR folder FOIA-85-158-A under your name.

The enclosed Appendix C further addresses documents pertaining to your request that could not be located because of insufficient identification or are not in the possession of NRC.

The denied portions of the records identified on Appendix B contain information i

which identifies procedures for safeguarding licensed special nuclear material

! at a licensed facility. This information is considered commercial or financial (proprietary) information pursuant to 10 CFR 2.790(d) and is being withheld from public disclosure pursuant to Exemption (4) of the F0IA (5 U.S.C.

552(b)(4)) and 10 CFR 9.5(a)(4) of the Connission's regulations.

Pursuant to 10 CFR 9.9 of the NRC's regulations, it has been determined that the information withheld is exempt from production or disclosure, and that its production or disclosure is contrary to the public interest. The persons responsible for-this denial are the undersigned and Mr. John G. Davis, Director, Office of Nuclear Material Safety and Safeguards.

-This denial may be appealed to the NRC's Executive Director for Operations

-within 30 days from the receipt of this letter. As provided in 10 CFR 9.11, any such appeal must be in writing, addressed to the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and should clearly state on the envelope and in the letter that it is an " Appeal from an Initial FOIA Decision."

8512160270'851205

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PDR FDIA RATNER 85-1 8 PDR

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f Mozart -G. ' Ratne'r, . Esqui re 1

We will communicate with you further regarding documents 1, 2, 6,13 and 24 of your request dated February 25, 1985. ~

Sincerely, M# 9

l. Donnie H. Grimsley, Director i Division of Rules and Records Office of Administration

Enclosures:

As stated i

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Re: F0IA-85-158-A (Second Resp'onse)

APPENDIX A NOTE: The number in parentheses following the description of the document corresponds to the number of the document listed in the request.

1. 10/13/84 Ltr from Vaughan to Crow, NMSS, (1.page) with attachment:

a. Attachment 1: Description of Revisions Made to 7/25/83 License Application (1 page)

b. Attachment 2: Revised License Application Pages dated 10/23/84 Revision 6-(38 page.s) (8)

NOTE: Regarding doc ment (8) as identified in your r'equest, the document cannot be identified as Section 2.9, dated 10/23/84. However, the suomittal dated 10/23/84 as listed above is being placed in the PDR. This record also addresses documents (9) (10) and (11) of your request.

2. 12/10/84 Ltr from Vaughan to Crow, NMSS, with following attachment:

a. 6/1/84 non-proprietary submittal for the UPMP project with enclosed Affidavit dated 6/4/84 (225 pages) -(16)

NOTE: The nonproprietary. portion of the 6/1/84 submittal was returned to the licensee at the licensee's request prior to receipt of F01A-85-158. The nonproprietary portion was resubmitted by General Electric on 12/10/84.

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Rei F01A-85-158-A (Second Response)

APPENDIX B PORTIONS OF RECORDS DENIED F0IA EXEMPTION (4)

NOTE: The number in parentheses following the description of the document corresponds to the number of the document listed in the F0IA request.

. 1. 10/12/84 Ltr f' rom Vaughan to Brown, subject: Exemption from Requirements.Regarding Transfer Document LE Calculations

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(1page) Released i

a.

Attachment:

10/12/84 Request for an exemption (1 page) -

Withheld - Exemption (4) (12)

2. 08/31/84 Ltr from Vaughan to Brown, subject: Request for One Time Exemption in Methodology for Reporting Inventory Values.

(1 page) Released

a.

Attachment:

8/31/84 Request for one time exemption (2 pages)

~ : Withheld - Exemption (4) (17).

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- Re: F01A-85-158-A

. . ' (Secon'd Response)

APPENDIX C NOTE; The document number corresponds to the number-in the request, Document Number Response

4. Insufficient identification. Request does not state year for 12/11 letter. If it is 12/11/84, then the information is available from the PDR in Docket 70-1113.

5. See number 4 above'.

20. NRC is not in possession of this document

- 22. NRC is not in possession of this document 9 -

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Washington, D. C. 20555 Director Division of Security U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. James M. Taylor Deputy Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D. C. 20555 REQUEST FOR PRODUCTION OF DOCUMENTS PURSUANT TO 10 CFR S 2.790(6) AND REQUEST FOR COMMISSION DECLASSIFICATION UNDER PART 9, APP. 2 (10 CFR 195-196)

Vera M. English, by her undersigned counsel, hereby requests that the enclosed list of documents, all of which are in possession of NRC but are not on file in the public document room, be made available for inspection and copying, without restriction, as soon as possible. Mrs. English is complainant in DOL Case No. 85-ERA-2, in which the Secretary of Labor, after investigation, has found reasonable cause to believe that G.E.'s highest management, in its Wilmington, North Carolina, Nuclear Manufacturing Plant, discriminatorily transferred and discharged. Mrs. English from her' analyst job in the Wet Lab, because1 she constantly -complaine'd to management, and finally to-

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NRCaboutnuchear.'safetyviolationsandhazards, quality ae ieth i A (

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con. trol deficiences and management's delibera'te fa'Isificatibn and cover up violations. If substantiated, the complaint of discrimination is a Severity Level I violation. (" Misc.

Matters," A, 4, 49 F.R. 8593)

The DOL hearing is scheduled to resume March 18, 1984 before ALJ Brissenden. NRC should recognizefan obligation to

~

produce, before the, reopening date, all releva'nt evidence in its possession bearing upon this violation. 10 CFR S 2.790(a)

(" violation of a license").

NRC regulations require the licensee to post, inter alia (10 CPR S 19.11(a) all "(2).* *

documents incorporated into a license by reference, and amendments thereto, (3) the ,

operating procedures applicable to licensed activities; (4) any notice of violation involving radiological working conditions

*

an order issued pursuant to Subpart B of Part 2 of this chapter and any response from the licensee" and (10 CFR S 21.6(a)(3)), all procedures adopted pursuant to regulations in this part.

In addition, NRC inspectors in Region II have conducted investigations and issued reports on Mrs. English's charges against G.E. Mrs. English is party to a proceeding against G.E. and the Second Region inspection staff of NRC, catagorized by Mr. Taylor as a 10 CFR S 2.202 proceeding

[7590-01], in which Mrs. English charges that the NRC inspectors' reports are deficient, inaccurate, biased and

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u unlawful - i .g. , violative of . NRC's published ' standards, by ,

which NRC is bound, but which the inspectors did not apply.

The undersigned received from NRC copies of various I reports, 82-18, 84-04, 84-05, 84-13, 84-15, 84-16, 84-17, ,

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. , _ . 84-18, under a protective agreement (10 CFR S 2.790(6)(i)),,

which Mr. Neal Abrams and M.r. Ed Shomaker stated was a condition precedent to. receipt of these reports.

The undersigned does not agree that the aforesaid reports and/or the documents listed in the enclosure may lawfully be " deemed to be commercial or financial in formation within the meaning of S 9.5(a)(4) of this chapter." The

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undersigned as'serts that as applied to a. litigant in such proceedings as detailed above, 10 CFR S 9.5(4), which authorizes withholding from public disclosure as

" confidential," matter "which is customarily held in confidence by the originator," is an unconstitutional denial of due process inasmuch as it denies complainant access to evidence necessary, or at least relevant, to prove her case and thereby vindicate her statutory right. It also frustrates performance by the charging party, "as private attorney general," of the role Congress assigned such parties "in enforcing the ban on discrimination." EEOC v. Associated Dry Goods Corp., 449 U.S.

590, 602 (1981).

Likewise, the undersigned asserts that subsection 4(i),

which exempts from disclosure "(i) [i] nformation received in I confidence, such as trade secrets, inventions.and discoveries l

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- a'nd' proprietary ' data," is ur7onstitutio'nally..overbroad insofar as it exempts "information received in confidence" and

" proprietary data," whatever that ambiguous, undefined, term may mean.

We have no quarrel with exemption from public disclosure of real " trade secrets, inventions' and discoveries" and material properly classified as " Safeguards .Information." But at maximum, only such portions of the inspect-lon reports and the documents enumerated on the enclosed list, which truly contain such information, may lawfully be withheld from public ex posu re . The " protective agreement" which the undersigned executed in exchange'for receiving the documents' applies to

, entire documents, not merely identified matter therein which can lawfully be withheld from public disclosure. To this extent, the " protective agreement" is legally overbroad and void.

S 10 CFR S 2.790 (b)(1), requires that "a person who proposes that a document or a part be withheld in whole or in part from public disclosure on the ground that it contains trade secrets or privileged or confidential commercial information" shall submit an affidavit requesting withholding which "may designate with appropriate markings the " trade secret or confidential or privileged commercial information within the meaning of S 9.5 (a)(4)," the objector desires withheld. Thus, the burden of proof is squarely placed on the

! objector to designate the parts of documents claimed to be F

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l exempt from public disclosure,under S 9.5(a)(4).- Absent a' swo'rn c'laim and proof by the opponent, G.E., that: portions of the documents requested fall within S 9.5(a)(4), NRC is not permitted to withhold any document from public inspection, except on " Safeguards Information" grounds. In this respect also, the " protective agreement" is overbroad and void.

Further, G.E. has waived any " confidential" -privilege it may have' claimed u'nder 10 CFR S 2.790, by providing, complainant in 85-ERA-2 with papers and data stamped " Company Confidential" and permitting complainant to offer those documents in evidence without objection or any request for in camera inspection or for a protective order. An example is Exhibit C-10, in.85-ERA-2, attached to this~ letter. Und.er the

" opened door" doctrine, that waiver extends not only to 1

documents named therein and to all like or related documents, but to all documents relevant to the charge in 85-ERA-2.

To the extent that the decisions referred t< and any document on the enclosed list of data may be withheld by NRC from public inspection, and receivable in evidence only subject to NRC's non-public access restrictions, we request that all _

restrictions be removed except from those portions of the documents as may be designated by G.E. which NRC, after careful review, determines are legally excludable. Of course, until NRC has released the documents for public inspection, or has been ordered by a court to do so, the undersigned, while at all

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't'imes reserving Mrs. Eng'lish's rights and claims in the matter, .

will abide b'y the " prot'ctive'a~greement."

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To the extent that any documents are withheld from

,public disclosure on the ground that they contain " Safeguards Information," we request " Declassification Review" under Part 9, App. A, pp. 195-196, 10 CFR (1/1/84 ed.). Accordingly, a copy of this request is also being submitted to the Director, Division of Security.

Because of the closeness of the trial date, the recent issuance of many of the subject reports (84-15 and 84-16 were received February 25, 1985, and 84-17 and 84-18 were received February 11, 1985), we respectfully urge consideration of and response to this " Request" on an emergency basis.

Very truly yours, Mozar G. Ratner Counsel for Vera M. English cc: James Lieberman, Esq.

,/ Chief Counsel V Regional Operations and Enforcement Neal Abrams, Esq., Senior Attorney Office of Executive Legal Director F. C. Shomaker, Esq.

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' REPORTS SUBMITTED. 8Y GENERAL E"LECTRIC COMP NY - .

(NUCLEAR ENERGY BUSINESS'0PERATIONS)~

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NUCLEAR REGULATORY COMMISSION - + - - - -

FOR VIOLATIONS OF NRC REGULATIONS DURING THE PERIOD -

JANUARY 1, 1978 TO July 27, 1984 NEfflAL VI0lATIONS JAIJUARY 9,1979 UF6 GAS RELEASE FETRUARY 14,1979 U02POWDERIHEFT I I#AY22,1900 '

UtMUTFORIZED RBC/AL OF lD2 PELLETS f l JUNE 4,19@l t

kCIDEtlTAL LOSS OF NASTE Llouro '

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i' ARCH 10,1979 UttAUT110RIZED ROU/AL OF C0tiTAMINATED IRASH JAtAJARY 1,1980 SHIPMEt(T OF UF6 CYUf0ERS t

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Charles M. Vaughan

] Manager, Regulatory Compliar.cc e

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LIST OF REQUESTED NRC DOCUMENTS NOT .ON FIL8 IN NRC PUBLIC DOCUMENT ROOM

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c ,qg;,g'/b*h 1. Letter of November 2, 1984, from Vaughan to Stohr're site

\ V s, l specific training program for trainees from G.E. to NRC. '

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) VV / 2. Letter of November 2, 1984, from G.E. to NRC describing _

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$ ,( training program, h r-w;j .

M. Attachment to November 15, 1984 Vaughan to O'Reilly letter M6, regarding results.of inspection report 84-11 (dated 10-18-84, on f.ile).

/ 4. Safety Evaluation.M'emo for Radiological' Contingency Plan',

Q identified in letter 12/11, Page to Vaughan (on file). ; n'd

. 5. Safety Evaluation Report for Changes to Chapter 4, ,

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identified in 12/11 letter Page to Vaughan (on file).

6. Letter from Vaughan to O'Reilly dated 2/14/84 4/; * -

7. Letter from Stohr to Long dated 8/10/84.

I (6 and 7 are referred to in 10/31/84 Vaughan to OReilly

! - letter in file)

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8 '. " Revised license application pages" - dated 10/23/84,

-Revision 6 submitted with letter dated 10/23/84 from

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Vaughan to W.T. Crow, NRC section leader; specifically, f,

Sec. 2.9 " Investigator's Report of Unusual Occurrence" -

/ I-2.20 and " Records" cec. 2.10. I-2.21 c 'G

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gfX 9. Ibid, Chapters 3 - Radiation Protection sec. 3. " Admin.

Regs" I-3.1; " Technical Regs" sec. 3.2; I-3.4; " Safety."

'N 10. Ibid, Chapter 9 - Overview of Operation.

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11. Ibid, Chapter 10, 11, 12, 13, 14, 15, 16.

N 12. Attachment to 10/12/84 Vaughan to Brown letter, on exemption regs, referred to in 9/28/84 Brown to Vaughan

'N f letter.

13. Material transmitted with letter from Chas. M. Vaughan to 1

J.P. O'Reilly dated March 21, 1984 (source Vaughan to Brown letter, 9/28/84, in file).

Letter from Stohr to Long dated August 10, 1984 (same 2

source as 13).

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! yg,1k,iS.G.E.'sRadiologicalContingencyPlansubmittedtoNRCon o -- January 14, 1982, and supplemented on April 4, 1984.

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i 1 (6/4/84) concerning UPMP h p/ h}6. Application (6/1/84)/

(referred to in 9/11/84Affidavit letter, Cunningham to Vaughan). i i

[ h pQ7. Fundamental Nuclear Material Control Plan submitted August g\ 31, 1984 (ref'd to in 9/7/84 letter from Brown to i Vaughan).

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18. Letters'of February 14, 1984 and March 21,.1984, relat'ing jd,')

to the Fundamental Nuclear Control Plan, from Long to Stohr

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referenced to in 8/10/84 letter from Stohr to Long.

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19. Enclosure to letter of August 10, 1984 from Stohr to Long

--_ " (FNCP).

N '20. P/P 40-17, Rev. 3, Nuclear Safety Traiing (referenced in 84/10 p, 3, 11 7 and 8.

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M . C . 41808 and M.C. 71814 (11 7 and 8, respectively, of fT,'" 84/10 Inspection Report).

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s . 2 2. PROD operating procedures ref'd at p. 4 - 84/10 1 8(d)

23. Pesponse of G.E. dated 6/7/84 to notices of violation

( issued 10/31/83 and 5/11/84 - referred to in letter from I

Stohr to Long dated 6/22/84 concerning Report ,

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570-1113/83-28 .

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, 24. All responses of G.E. to findings of violation between 1978 i -

and 1985. This is a continuing request.

25. Letter from J. O. O'Reilley to J.A. Long dated July 30,

- 1984, summarizing meeting held in Atlanta (II) on July 10, 1984, between G.E. Reps and O'Reilly et al., ref'd in letter from Vaughan to O'Reilly dated 8/29/84.

N26. Attachment to June 7, 1984 letter from Vaughan to Stohr (responses to findings of " referenced" inspection NRC Insp.

Rept. 83-28 (10/31/83).

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9 February 22, 1985 I, Vera M. English, authorize my attorney, Mozart G.

Ratner, to receive all correspondence, transcripts, and other

' documents pertaining to " Request for Action Under 10 CFR 2.206 Regarding Activities at

i the Wilmington, North Carolina Facility d

of-the General Electric Company," Docket NO. 70-1113, and English v. General Electric Company, DOL Case No. 85-ERA-2.

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.ex v" a-GENERAL @ ELECTRIC W!t.MINGTON MANUTACTUR:NG DEPARTMENT GENERAL ELECTR!C COM%NY + P.O. BOX 780 VAIMINGTON, NORTH CAROUNA 28402 Octcber 12, 1984 ,_.

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Director {g $ v Office of Nuclear Materials Safety & Safeguards -

U. S. Nuclear Regulatory Commission Washington, D. C. 20555

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Attention: Mr. W. B. Brown, Chief o;

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Fuel Facilities Safeguards Licensing Branch ,

Dear Sir:

Reference:

NRCLicenseSNM-1097,bocket 70-1113

. S& ject: EXEMPTION FROM REQUIREMENTS REGARDING TRANSFER DOCUMENT LE CALCULATIONS With reference to activities authorized 'by NRC License SNM-1397 at the General Electric fuel fabrication plant in Wilmington, North Carolina, GE hereby submits a request for an exemption concerning limit of error calc 41ations as delineated in the attachment to this letter.

Pursuant to 10 CFR 2.790(d), GE requests that this attachment be withheld from public disclosure as it contains details of General Electric procedures for safeguarding licensed special nuclear material.

Pursuant to 10 CFR 170.31, a check for $150 is enclosed for processing thf: request.

General Electric personnel would be pleased to discuss this matter further with you and your staff as you may Jecm necessary.

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- Very truly yours,

,gre GENERAL ELECTRIC COMPANY

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As g gn Wr e .. fs, Charles M. Vaughan, Manager Licensing & Nuclear Materials Management

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.. ". *a ?.. F~" W..O. "t '>.ra 1 I':I F.5THODOLOG'l FOR RZPORTING INVENTORY VALUES nith W farence to activiti.?s authorized by NRC License SMM ~1097 at the C2 fugl ra.bricctica plant .in tiilming ton, North Carcl'ina ,

riener 41 21ect.ric hereby reques ts . authorization, . cn a one time basi:, use 1985 measur:ad valueq -in lieu of 19p4 naasured ve.lueu. :n the statcuent of a portion of the current inventory as cielineated in the attachment to this letter.

Mrsutut to 10 CFR 2.790f.d), General Electric req.:ests that the ct.cac;:r.eni. to this 1.att.er be withhold frca public disclO;ury as it

'ec :tcin - infor::tirn tihi-h i.dsntifies General Slectric proc duras

27. sa fq erding lbzensed special nuclear material.

Purzudnt to !O-CTR 170.31. cnr:)ncud is a C-E check foe 5150 for i u,r<:.m a a i ns this ste:ndnu t request.

l Gensrni ilnctric personnel would be pleased to discus; .this matter t

further with you ar.d your staf f as you may .ies, necassary. ,,

very truly yours, i

GENERAL ELECTRIC COM~4NY -

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, GENERAL ELECTRIC COMPANY

PO BOX 780.Wl'MNGTON, NORTH CAQiNA 98402 October 23, 1984- , , , ,

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U. S. Nuclear Regulatory Commission Washington, D. C. 20555 /- g W;W ,r- ..., -3, I.: < C:.T

2 .- 5 Attention: Mr. W. T. Crow, Section Leader l- O Uranium ~ Process Licensing Section \(,' g, at. :. ,, f % ' *c"E.b M/S 396-SS. ,,.,,.;

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Dear Sir:

References:

NRC License SNM-1097, Docket #70-1114

Subject:

LICENSE A'iLNDP.EN' REQUEST - REVISION 16 s General Electric Company Nuclear Fuel Manufacturing Department hereby requests revision of License Condition 9 of NRC License SNM-1097 'to incorporate the attached revised pages of Part I of the GE application dated 5/14/84, as amended.

Attachment 1 details the revisions madd and Attachment 2 contiains the revised pages. '

Pursuant to 10 CFE'170.31, on 10/4/84 a. check for $150 was i submitted to the NRC for processing an exemption request. On 10/23/84, this req'uest was withdrawn. By. copy of this letter, the NRC License Fee Management Branch is requested to P.ransfer the S150 credit to.thi license amendment request.

General Electric per'sonnel would be pleased to discuss these matters further,with you and your staff as deemed necessary.

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N; } Very truly yours,

/,1 - : GENERAL ELECTRIC COMPANY g;W['I ,. ,

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I f r GENERAL h ELECTRIC Director - ONMSS October 23, 1984 ATTACHMENT 1 DESCRIPTION OF REVISIONS MADE TO 7/25/83 LICENSE APPLICATION Page Description 1 thru Table of Contents page numbers were changed from Roman 12 numerals to numerals. Pages 7-12 were updated and reformatted to reflect Part I pages revised on 10/23/84.

-I-1.1 Revised to reflect a change in name for the Wilmington facility from Wilmington Manufacturing Department to Nuclear Fuel Manufacturing Department, effective 10/16/84. .

Chapter 4 This chapter has been revised to include the use of concentration and density controls and to more

, accurately reflect the way in which limits and

, _ controls are applied at the Wilmington facility. The basic philosphy under which NFMD operates has not been changed; criticality safety is still required to comply with the requirements of Section 4.1.1.

Section 4.2.1 has been added to reflect specifically licensed limits for geometry limited systems, for mass limited systems, and for concentration limited systems.

Sections 4.2.2 through 4.2.12 have been revised to include for each of the parameters used in criticality ,

safety, the requirements which must be met in order to use values which have not been specifically licensed.

1 These modifications to Chapter 4 are not only in

support of the Uranium Process Management Project i amendment request currently under NRC review, but will also facilitate future licensing actions in other' plant areas.

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GENERAL $ ELECTRIC Director - ONMSS October 23, 1984 4

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ATTACHMENT 2 REVISED LICENSE APPLICATION PAGES DATED 10/23/84 6 REVISION 6 1

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3 ) TABLE OF CONTENTS Section Title Page e

PART I - LICENSE CONDITIONS CHAPTER 1 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS 1.1 Corporate & Financial Information ............. I-1.1 1.2 Location & General Description of Wilmington Plant ......................................... I-1.1 1.3 License Number ................................ I-1.1 1.4 Po s s e s s ion L im i t s . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . I- 1. 2

, 1.5 Material Use Locations ........................ I-1.3 1.6 Definitions ................................... I-1.3 1.7 Authorized Activities ......................... I-1.5 g 1.8 Exemptions & Special Authorizations ........... I-1.8

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CHAPTER 2 - GENERAL ORGANIZATION & ADMINISTRATIVE REQUIREMENTS 2.1 Policy ........................................ I-2,1 2.2 Organizational Responsibility & Authority ..... I-2.1 2.3 S a f e ty Re'r iew Commit t ee . . . . . . . . . . . . . . . . . . . . . . . I-2. 6 2.4 Approval Authority for Personnel Selections ... I-2.8 2.5 Personnel Education & Experience Requirements . I-2.9 2.6 Training ...................................... I-2.13 2.7 Operating Procedures - Administrative Controls I-2.14 2.8 Audits & Inspections .......................... I-2.17 l

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2.9 Investigation & Reporting of Unusual Occurrences ................................... I-2.20 2.10 Records ....................................... I-2.21 CHAPTER 3 - RADIATION PROTECTION 3.1 Administrative Requirements ................... I-3.1 3.2 Technical Requirements ........................ I-3.4 CHAPTER 4 - NUCLEAR CRITICALITY SAFETY 4.1 Administrative Requirements ................... I-4.1 4.2 Technical Requirements ........................ I-4.4 CHAPTER 5 - ENVIRONMENTAL PROTECTION 5.1 Effluent Control Systems ...................... I-5.1 5.2 Environmental Monitoring Program . . . . . . . . . . . . . . I-5.13 CHAPTER 6 - SPECIAL PROCESS COMMITMENTS 6.1 UF 6 Cylinder Movement ......................... I-6.1 6.2 00 2 Powder Moisture Analysis & Moisture Limit . I-6.1 6.3 Transfer of UPS & Fluoride Liquid Wastes from Safe to Unsafe Geometries ................ I-6.1 CHAPTER 7 - DECOMMISSIONING PLAN ....... I-7.1 CHAPTER 8 - EMERGENCY PLAN .......... I-8.*

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Section Title Page I

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PART II - SAFETY DEMONSTRATION '

CHAPTER 9 - OVERVIEW OF OPERATION 9.1 Corporate Information ......................... II-9.1 9.2 Corporate Financial Qualification ............. II-9.1 9.3 Summary of Operating Objective & Process ...... II-9.1 9.4 S ite De s cript ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . II-9 . 2 9.5 Location of Build ings On S ite . . . . . . . . . . . . . . . . . II-9.2 9.6 Location of the Plant Site .................... II-9.2 9.7 License History ............................... II-9.2 9.8 Significant Changes in Corporate Structure . . . . II-9.8 9.9 Changes in Procedures, Facilities & Equipment . II-9.8 CHAPTER 10 - FACILITIES DESCRIPTION 10.1 Plant Layout .................................. II-10.1 10.2 Utilities ..................................... II-10.6 10.3 Heating, Ventilation & Air Conditioning (HVAC). II-10.13 10.4 Waste Handling ................................ II-10.26 10.5 Chemical Systems .............................. II-10.32 10.6 Fire Protection ............................... II-10.34 CHAPTER 11 - ORGANIZATION & PERSONNEL 11.1 Section Functions ............................. II-11.1

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. j i ) TABLE OF CONTENTS Section Title Page 11.2 Organization Charts ........................... II-11.3 l l

11.3 Organizational Procedures ..................... II-11.3 I 11.4 Key Functions ................................. II-11.9 11.5 Education & Experience of Key Personnel ....... II-11.12 11.6 Training ...................................... II-11.53 CHAPTER 12 - RADIATION PROTECTION PROCEDURES & EQUIPMENT 12.1 Procedures for Radiation Surveys .............. II-12.1 12.2 Posting & Labeling ............................ II-12.2 12.3 Personnel Monitoring .......................... II-12.3 12.4 Operational Surveys ........................... II-12.6 12.5 Records ....................................... II-12.7

( 12.6 Reports ....................................... II-12.8 12.7 Instrumentation ............................... II-12.9 12.8 Protective Clothing ........................... II-12.15 12.9 Administrative Action Guidelines .............. II-12.15 12.10 Respiratory Protection ........................ II-12.17 CHAPTER 13 - OCCUPATIONAL RADIATION EXPOSURE

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13.1 Occupational Exposure' Analysis ................ II-13.1 13.2 ALARA at WMD .................................. II-13.54 13.3 Bioassay Program .............................. II-13.67 l

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$ Section Title Page 13.4 Air Sampling & Assigned Airborne Exposure ..... II-13.73 13.5 Surface Contamination Control Program ......... II-13.80 13.6 Shipping & Receiving Controls ................. II-13.85 CHAPTER 14 - ENVIRONMENTAL SAFETY 14.1 Radiological .................................. II-14.1 14.2 Non-Radiological .............................. II-14.1 CHAPTER 15 - NUCLEAR CRITICALITY SAFETY 15.1 Administrative Practices - Area Managers ...... II-15.1 15.2 Administrative Practices - Criticality Safety Function ............................... II-15.7 15.3 Equipment & Facility Design Considerations .... II-15.10

(~- 15.4 Methodology in Criticality Safety Analyses .... II-15.14 15.5 Analytical Methods ............................ II-15.19 15.6 Method Validation ............................. II-15.29 CHAPTER 16 - PROCESS DESCRIPTION

& SAFETY ANALYSES 16.1 Receiving & Storing UF 6 ....................... II-16.1 16.2 UF6 -to-UO2 - Conversion - ADU Process . . . . . . . . . . . II-16.6 16.3 UF S -to UO 2 Conversicu - GECO Process .......... 11-16.18 16.4 00 2 Powder Pre-Treatment ...................... II-16.28 LICENSE SNM-1097 DATE 10/23/84 -P AG E DOCKET 70-1113 REVISION 6 l .

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16.5 00 2 Powder Blending ........................... II-16.36 16.6 Pellet Production ............................. II-16.38 16.7 Sintering ..................................... II-16.40 16.8 Pellet Grinding ............................... II-16.41 16.9 Fuel Rod Loading .............................. II-16.42 16.10 Fuel Bundle Assembly .......................... II-16.44 16.11 Fuel Bundle Leak Test & Final Inspection ...... II-16.47 16.12 Fuel B undle S torage . . . . .'. . . . . . . . . . . . . . . . . . . . . . II-16. 4 8 16.13 Packaging of Fuel Bundles for Transport ....... II-16.50 16.14 Scrap Recovery ................................ II-16.51 16.15 Waste Treatment & Disposal .................... II-16.55 16.16 Chemical-Metallurgical Laboratory ............. II-16.65 16.17 Outside Product Can Storage ................... II-16.67 16.18 Process Technology Laboratory ................. II-16.68 CHAPTER 17 - ENVIRONMENTAL EFFECTS OF ACCIDENTS ................. II-17.1

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I-6.1 6/20/84 I-4.23 I-4.24 I-4.25

CHAPTER 7 I-7.1 5/14/84 CHAPTER 5 i I-5.1 5/14/84 I-5.2 CHAPTER 8 I-5.3 I-5.4 I-8.1 9/24/84 I-5.5 I-5.6

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( I-5.8 6/20/84 I-5.9 5/14/84 I-5.10 I-5.11 I-5.12 I-5.13 I-5.14 6/20/84 I-5.15 5/14/84 I-5.16 6/20/84 I-5.17 5/14/84 I-5.18 6/20/84 I-5.19 I-5.20 5/14/84 I-5.21 6/20/84 I-5.22 I-5.23 I-5.24 I-5.25 I-5.26 I-5.27 I-5.28

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REVISIONS BY PAGE Effective Effective Effective Page Date Page Date Page Date II-10.22 7/25/83 II-11.15 7/25/83 PART II II-10.23 II-11.16 II-10.24 II-11.17 II-10.25 II-11.18 CHAPTER 9 .II-10.26 II-11.19 "

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II-9.3 II-10.30 II-11.23 II-9.4 II-10.31I II-11.24 II-9.5 II-10.32 II-11.25 II-9.6 II-10.33 II-11.26 II-9.7 II-10.34 II-11.27 II-9.8 II-10.35 II-11.28 "

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CHAPTER 1 g

STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1.1 CORPORATE & FINANCIAL INFORMATION This licensing information document is filed by the Nuclear Fuel Manufacturing Department of the General Electric Company, a New York corporation with a principal place of business at Schenectady, New York, i

1.2 LOCATION & GENERAL DESCRIPTION OF WILMINGTON PLANT The General Electric Company, Nuclear Fuel Manufacturing

Department (NFMD) operates a nuclear fuel fabrication

plant in Wilmington, North Carolina. At this site, NFMD

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occupies buildings for administrative, laboratory and manufacturing activities. Fuel manufacturing activities are conducted within the Fuel Manufacturing area.

The full address is as follows: General Electric Company, Nuclear Fuel Manufacturing Department, (name of

person and mail code), P. O. Box 780, Wilmington, NC 28402.

1.3 LICENSE NUMBER The General Electric Company Nuclear FJel Manufacturing

Department NRC license number is SNM-1097 (Docket 70-1113).

k LICENSC SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVIS IO!! 6 I-1.1 9

l CHAPTER 4 NUCLEAR CRITICALITY SAFETY l 4.1 Administrative Recuirements 4.1.1 Process / Facility Design Philosophy Process designs shall incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is' possible.

The preferred method for assuring nuclear criticality safety in production quantities of fissile material is by the use of safe geometry. However, other controls

may be used when safe geometry is not practical. For *

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example, batch control may be used in situations involving experimental quantities of material of from 7%

to 15% enrichment. (See Table 4.4, Page I-4.9)

Fixed neutron absorbers may be used as part of a safe

geometry. Such use is preferred over the use of

administrative controls.

The use of administrative controls for nuclear criticality safety will be restricted to those cases in which safe geometry is not practical.

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4.1.2 Recuests for Nuclear Criticality Analyses The Area Manager assures that all changes, modifications or additions to the plant which have the potential to affect criticality safety, are appropriately reviewed and approved for criticality safety prior to implementation.

If the Area Manager concludes that a proposed new activity or change in activity has the potential to affect criticality safety, a request for a nuclear criticality analysis shall be submitted in writing to the criticality safety function. After review of tha

, request, a member of the criticality safety function having the qualifications of a senior member as described in Section 2.5.2.1, shall make the decision as to whether or not an analysis is required.

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If an analysis is required, the change shall not be placed into operation until the criticality safety analynis is complete and the concomitar.t nuclear criticality safety requirements and approval for operation are received from the criticality safety function.

4.1.3 Criticality Control Procedures Each Area Manager shall assure that criticality control procedures incorporating limitations established by the criticality safety function are developed and maintained and shall assure that foremen, operators and other concerned personnel are made aware of these procedures

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. I 7 'through posting, training programs or other appropriate

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written notifications.

j- 4.1.4 Posting of Nuclear Criticality Safety Limits 1

Nuclear criticality safety requirements which are j received from the' criticality safety function for each process system shall be available at each work station either in the, form of operating procedures or as clear, visible signs or notices.

Posted nuclear criticality safety requirements shall be defined by the criticality safety function and may include: limits on material types and forms; allowable quantities by weight or number; allowable enrichments; required spacings between units; control limits, when applicable, on quantities such as moderation, and I(

density or the presence of additives.

4.1.5 Labeling of Containers of Fissile Material e

l Containers of fissile material (not including fuel rods, shipping containers, waste boxes / drums, samples and the 4

like) shall be labeled such that the material type, U 235

!. enrichment and gross and net weights can be clearly

identified or determined.

4.1.6 Role of the Criticality Safety Function Personnel of the criticality safety function determine safe batches,; safe geometries, safe concentrations, and safe spacing of special nuclear materials; they shall i

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, determine limitations of other nuclear parameters utilizing information set forth in the request for criticality analysis; and they assess and verify the normal and potential environmental conditions of significance to criticality control.

Criticality safety function personnel also perform audits and provide advice on training.

4.2 TECHNICAL REQUIREMENTS 4.2.1 Specific License Limits

4.2.1.1 The safe geometry values of Tables 4.1, 4.2 and 4.3 are

specifically licensed for use in the Wilmington

facility. Application of these geometries is limited to

situations where the neutron reflection present does not *

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exceed that due to full water reflection.

When cylinders and slabs are not infinite in extent, the

dimensional limitations of Tables 4.1, 4.2, or 4.3 may

be increased by means of standard buckling conversion

methods or reactivity formula calculations which

incorporate validated k-infinites, migration areas (M2)

and extrapolation distances.

4.2.1.2 The safe batch values of Table 4-4 are specifically

licensed for use in the Wilmington facility.

Criticality safety may be based on U235 mass limits in

either of the following ways:

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I.4,4

TABLE 4.1 SAFE GEOMETRY VALUES FOR HOMOGENEOUS UO,-H,0 MIXTURES I Nominal Infinite Infinite Weight Cylinders Slabs Spheres Percent Diameter, Thickness, Volume, U235 Inches Inches Liters a

2.00 16.7 8.9 105 2.25 14.9 7.9 75.5 2.50 13.75 7.2 61 2.75 12.9 6.65 51 3.00 12.35 6.25 44

,- 3.25 11.7 5.9 38.5

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3.50 11.2 5.6 34 3.75 10.8 5.3 31 4.00 10.5 5.1 29 5.00 9.5 4.45 24 6.00 8.95 4.00 18.5 7.00 8.45 3.75 17.0 I For enrichment not specified in this table, smooth curve interpolation may be used.

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C TABLE 4.2 SAFE GEOMETRY VALUES FOR HOMOGENEOUS AQUEOUS SOLUTIONSI Infinite Nominal Slabs Spheres W/O Thickness, Volume, U235 Inches Liters 2.00 9.3 106.4

', 2.25 8.4 80.5 2.50 7.8 66.8 2.75 7.3 56.2 3.00 7.0 49.7 3.25 6.7 44.8

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3.50 6.5 41.0 3.75 6.3 38.0 4.00 6.0 34.9 5.00 4.8 26.0 6.00 4.4 22.5 7.00 4.1 19.5 I

For enrichments not specified in this table, smooth curve interpolation of values may be used.

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(

TABLE 4.3 SAFE GEOMETRY VALUES FOR HETEROGENEOUS MIXTURES OR COMPOUNDS I Infinite Infinite Nominal Cylinders Slabs Spheres W/O Diameter Thickness, Volume, U235 Inches Inches Liters 2.00 11.1 5.6 35.7 2.25 10.5 5.1 30.7 2.50 10.1 4.8 27.3 2.75 9.7 4.6 24.7 3.00 9.4 4.4 22.6 3.25 9.2 4.3 20.9

(-- 3.50 9.0 4.2 19.2 3.75 8.9 4.1 18.2 4.00 8.8 4.0 16.9 5.00 8.3 3.6 13.0 6.00 7.9 3.5 11.0 7.00 7.4 -

7.6 1

For enrichment not specified in this table, smooth curve interpolation of values may be used.

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 7_4,7 O

If double batching is credible, the mass of any

single accumulation shall not exceed a safe batch,

which is defined to be 45% of the minimum critical

mass. Table 4.4 lists safe batch limits for

homogeneous mixtures of UO 2 and water as a function

of U235 enrichment over the range of 1.1% to 15.0%

for uncontrolled geometric configurations.

Where engineered controls prevent cver batching, a

mass of 75% of a minimum critical mass shall not be

exceeded.

The safe batch sizes fo'r UO 2 of specific enrichments set

forth in Table 4.4 shall be adjusted when applied to

other compounds by the formula:

kgs 00 2 x 88% = kg X x f

where f = % U in Compound X

ls

4.2.1.3 Subject to provision for adequate protection against

precipitation or other circumstances which may increase

concentration, the following safe concentrations are

specifically licensed for use at the Wilmington

facility.

A concentration of less than one-half of the *

, minimum critical.

A system in which the hydrogen to U 235 atomic ratio

is not less than 5200.

1 N '

SNM-1097 DATE 10/23/84 PAGE LICENSC _

DOCKET 70-1113 RCVISION 6 I-4.8

TABLE 4.4 SAFE BATCH LIMITS FOR UO, & H,0 (Kgs 002)

Nominal U235 U235 Enrich- Enrich-ment UO 2 UO 2 ment UO 2 UO 2 w/o Powderl Pellets 2 w/o Powderl Pellets 2 1.1 2629 510 3.4 34.6 31.0 1.2 1391 341 3.6 31.1 28.5 1.3 833 246 1 3.8 28.3 26.4

. 1.4 583 193 4.0 25.7 24.7

. 1.5 404 158 4.2 23.7 22.9 1.6 293.3 135 4.4 21.9 21.4 1.7 225.0 116 '4.6 20.2 20.0 1.8 183.0 102 4.8 19.1 18.8 1.9 150.6 90.5 5.0 18.1 18.1

( 2.0 2.1 127.5 109.2 81.6 73.1 5.5 15.4 15.4 6.0 13.8 13.8 2.2 96.8 66.4 7.0 8.3 8.3 2.3 84.3 61.0 8.0 6.9 6.9 2.4 74.7 56.1 9.0 5.9 5.9 2.5 68.9 52.1 10 5.1 5.1 2.6 60.5 48.8 11 4.4 4.4 2.7 56.6 45.4 12 3.9 3.9 2.8 52.2 42.9 13 3.5 3.5 2.9 47.6 40.1 14 3.3 3.3

, _3.0 44.5 38.1 15 3.0 3.0 ,

3.2 38.9 34.1 NOTE: For enrichments not specified above, smooth curve interpolation of safe batch values may be used.

1 Homogeneous mixtures 2 Hetergeneous mixtures d.

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I_4,9

~. .

4.2.2 Nuclear Criticality Safety Methodology

4.2.2.1 Nuclear criticality analyses shall utilize experimental

data or analytical methods which have been bench marked by comparison with experimental data. An analytical method shall be considered bench marked when the following are established: (a) the type of systems which can be modeled, (b) the range of parameters which may be treated,, and (3) the bias, if any, which exists in the results produced by the method.

4.2.2.2 Each nuclear criticality analysis shall be verified by

an independent and qual'ified member of the criticality

, safety function having the qualificatons of a senior member as defined in Section 2.5.2.1 of this application. ,

4.2.2.3 When analytical methods are used in nuclear criticality *

( analyses to determine system neutron multiplication factors, it is required that at the 99 percent confidence level the neutron multiplication factor for normal operations be no greater than 0.90 and the neutron multiplication factor for all conditions required to be critically safe by the double-contingency policy be no greater than 0.97.

These restrictions shall be applied as follows:

K,+ 30 - bias 1 90 for normal conditions K, + 3o - bias 1 97 for failure of a single contingency Where K,= The average effective multiplication factor calculated for the desired application LICENSC SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISIO;l 6 I-4,10

}D

a = The statistical or convergence uncertainties in the calculation of K 3, and Bias = The bias of the calculational technique defined as the calculated result minus the

experimental result.

4.2.2.4 Criticality safety analyses shall take the following

into consideration:

Normal conditions - The most reactive values not

excluded by referenceable controls or system parameter

shall be assumed for the following variables:

Moderation

Reflection

i

(

Heterogeneity of fuel, absorbers, moderators and

structures

Possible buildup of fissile material in

inaccessible or unplanned locations

Interaction between individual pieces of equipment

Interspersed moderation between individual pieces

of equipment

Absorption characteristics of materials of

construction *

=.

LICENSC SNM-1097 DATE 10/23/84 PAGE 00CKCT 70-1113 RCVISION 6 I_4,11

5, . .-

g Enrichment

e Accident conditions - In accident condition

calculations, the controls cited in normal condition

calculations are assumed to fail in accord with the

requirements of Section 4.1.1 and the reactivity of the

system is reevaluated.

4.2.2.5 . Calculational, methods used in nuclear criticality safety

analyses shall include GEKENO and GEMER Monte Carlo Codes, keff calculations using the reactivity formula, and the solid angle technique of array evaluations.

Newly' developed codes o'r techniques may also be used

', when they have been validated and bench marked in accord with General Electric internal procedures and with the applicable ANSI standards.

- 4.2.3 Control Philosophy

4.2.3.1 A criticality safety control must be capable of

preventing a critical accident independent of the

i operation or failure of any other control

4.2.3.2 A single criticality safety control may be composed of *

. limits applied to two or more cystem parameters.

Controls on geometry, mass, moderation, concentration, *

. .- enrichment. and density may be used in such combinations. *-

For example, the volume of a system may be used in

combination with a concentration control to establish

l mass as a control for the system. * !

I LICENSE SNM-1097, DATE 10/23/84 PAGE coCKET 70-1113 REVIston' 6 I-4.12

4.2.4 Nuclear Criticality Safety Design Considerations for

Geometry Control

4.2.4.1 The use of geometry control in criticality safety,

except as specifically licensed in Section 4.2.1, shall

comply with the requirements of Sections 4.1.1, 4.2.2.3 and/4.2.2.4.-

4.2.4.2 Equipment used, to process or store fissile material and

designed on thu basis'of neutron multiplication is

considered to be s6fe geometry if, under normal

conditions, the fully reflected effactive multiplication * .

satisfies the normal co'ndition specifications of 4.2.2.3

and if, when accident conditions are credible, it also

satisfies the accident condition specifications of

Section 4.2.2.'O. *

~

4.2.4.3 Whenever c.riticality control is directly dependent on *

( -

the integrity of a structure used to retain the geometric forn of a fissile material accumulation or the spacing within a storage' array, the structure shall be designed wi.th an adequate strength factor to assure against failure under foreseeabic loads or accident conditions. Materials of construction shall be fire resistant. The degree to which any corrosive I environment might affect' nuclear safety shall be

. - . considered and corrosion-resistant materials or coatings-applied as necessary.

4.2.4.4 The use of fixed neutron absorber systems as geometric

controls shall require that: 3 f '

L IC P.!!S C S!!M-1097 DATE - 10/23/84 PAGS i

70CKE7 .74 1113 'R CV I S I O'!! . 6 I-4.13

1) The neutron absorber is one of the following:

a) Elemental cadmium b) Elemental boron alloyed with steel c) Solid, stable boron compounds such as boron carbide f!.xed in a matrix such as aluminum or polyester resin.

2) The neutron absorber and any hydrogeneous material used to thermalize neutrons are sealed in a stainless steel or other suitable container which may be an integral part of the process equipment.

3) Theneutronabsor$ersystem(i.e., absorber,

. moderator, container) is installed as a permanent part of the process equipment such that it cannot be readily removed.

4) Prior to using a fixed neutron absorber system for

(

criticality control, an inspection shall be performed using written procedures developed and approved by management, to verify the presence of the neutron absorbers and to verify that the system installation is in accordance with design and nuclear criticality safety requirements.

Inspection records shall be documented and maintained for the life of the system.

5) The void volume of the neutron absorber system is negligibly small to prevent internal rearrangement.

f LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.14

6) The effectiveness of the neutron absorber system is demonstrated utilizing validated calculational methods.

7) The integrity of the fixed neutron absorber system must be maintained against any credible fire hazard, i

8) The integrity of the fixed neutron absorber system must be verified on a periodic schedule compatible with the rates of corrosion and deterioration credible to the process system.

f

9) If the absorber is removed, the system shall be locked out in accordance with safety lock and tag procedures (e.g., circuit switch locked, circuit breaker removed, feed lines removed) ,or operated under controls authorized in writing bh the

(- criticality safety function which ensure, as a minimum, that the requiremento of Section 4.1.t' are

met. ,

10) The fixed neutron absorber system must be designed to withstand all cred'ible induF',tial accidents and natural events. ,

,. ..,,. 4 2.4.5

, Whenever criticality control.. is directly. dependent on .

the integrity of physical barriers or neutron absorbers, the structure shall be designed to assure against loss of' integrity through foreseeable accident conditions such as fire, impact, melting, corrosion or leakage of materials. .

\

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 ! I-4.15 1, ,_ _ _ _ .

, 4.2.4.6 Where control of the spacing and/or height of movable l s

units is used to provide criticality safety, the i geometry of the system is administratively controlled by * '

one or more of the following techniques:

The geometry of the system is defined by engineered

devices that determine the location of the movable *

! units. l i The geometry of the system is defined by procedure

through the use of visually identified storage 2

areas.

The geometry of the system is defined by engineered

devices that limit the height of fissile material.

i The geometry of the system is defined by procedure

through the use of maximum height restrictions.

(- ,

Where this type of control is used, movable units must

be safe geometry and must contain less than a minimum

i.

critical mass based on the form of the fissile material (i.e. , powder, pellets or rods) .

Nuclear Criticality Safety Considerations for

4.2.5 Administrative Control of Mass

- ., - 1 .

. .. .c ..., c ...:.. .. . ,

Where. control of mass is used to provide criticality

4.2.5.1 safety, the mass of uranium (or U 235 or U238) is

administratively controlled based on measurement by one

or more of the following techniques:

l'

' =i '

LICENSC SNM-1097 DATE 10/23/84 PAGE l DOCKET 70-1113 REVISION' 6- I-4.16 l b l

The mass of uranium (or U235 or U238) is determined

as the product of the volume and the uranium (or

U235 or U238) concentration as measured by

qualified counting methods.

The mass of uranium (or U23S OR U238) is determined

by qualified counting methods.

The tota.1 mass or change in nass of a system is

measured assuming the most reactive credible

composition.

4.2.5.2 The use of mass control in criticality safety except as

specifically licensed in Section 4.2.1, shall comply

with the requirements of Sections 4.1.1, 4.2.2.3 and

4.2.2.4.

- 4.2.6 Nuclear Criticality Safety Considerations for

Administrative Control of Moderation 4.2.6.1 Criticality safety of vessels, structures or processes

may be based on control of moderation provided that the following conditions are satisfied:

Sources of moderation internal and external to the

process shall be identified and controls

established for.each source which are consistent

with the requirements of Section 4.1.1.

Support equipment associated with the control or

processing of moderating materials shall be

designed so that they are either geometrically safe

or designed to prevent backflow of fissile

l=(( SNM-1097 DATE 10/23/84 PAGE l LICENSE DOCKET 70-1113 REVISION. 6 y_4,17

materials and/or flooding of the fissile

materials.

4.2.6.2 Where control of moderation is used to provide

criticality safety, the degree of moderation is limited

by one or more of the following techniques:

Moderation is removed by an engineered system and

the resulting material is inspected and sampled to

ensure proper functioning of the equipment.

Moderation is added procedurally and controlled by

limiting the mass and/or volume of the moderator *

. and fissile material.

Moderation of the mixture is determined by analysis

or engineered methods prior to using moderation as *

, a criticality control.

4.2.6.3 The control of moderation for purposes of criticality

safety must comply with the requirements of Sections

4.1.1, 4.2.2.3 and 4.2.2.4.

4.2.7 Criticality Safety Requirements for Use of Concentration

Control

4.2.7.1 Where control of uranium concentration is used to

provide criticality safety, the concentration is

controlled by one or more of the following techniques:

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.18

. l 1

~

e

. c The solubility is controlled to prevent 1 y~

precipitation and the material is either agitated

or recirculated at a rate sufficient to prevent *

! settling into an unsafe concentration.

The uranium concentration is limited by on-line

measurement of concentration (or density if the

worst credible composition is assumed). If the limit is reached, automatic controls must prevent 4

continued increase.

The uranium concentration in a precipitate is

measured and limit!ed assuming the worst credible

composition.

4.2.7.2 A full density mixture is used in determinations of

^

uranium concentraiton (i.e., the effect of voids or

inert materials mixed with the accumulation is not

included).

4 J

The control of concentration for purposes of criticality

4.2.7.3 safety, except as specifically licensed in Section 4.2.1, shall comply with the requirements of Sections

4.1.1, 4.2.2.3 and 4.2.2.4.

Criticality Safety Recuirements for Use of Density

4.2.8 i

-~ -

Control , , ,

Where control of uranium density is used to provide

4.2.8.1

=

criticality safety, the uranium density is administratively controlled by one or more of the following techniques:

l I:

LICENSC SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION. 6 I-4.19 l

l .

Fuel displacing media is maintained within the

geometry by use of mechanical attachments or by use

of a size / shape that cannot exit from the geometry

during use. Where this method of criticality

control is used, it is necessary to assume the

worst credible distribution of media within the

geometry.

The uranium density is controlled by use of

homogenized mixtures with measured uranium

density.

Where materials having different uranium densities are

mixed, the maximum density measured is assumed

throughout the system.

4.2.8.2 The control of density for purposes of criticality *

, safety must comply with the requirements of Sections *

k. 4.1.1, 4.2.2.3 and 4.2.2.4.

4.2.9 Criticality Safety Requirements for Use of Enrichment

Control

4.2.9.1 Where control of uranium enrichment is used to provide

criticality safety, the uranium-235 enrichment is

administratively controlled based on measurement by

standard assay techniques prior to'the enrichment being

used as a criticality control. Where material having

different enrichment values are mixed, the maximum

enrichment measured is assumed throughout the system.

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.20

4.2.9.2 The control of enrichment for purposes of criticality

'- safety must comply with the requirements of Sections

4.1.1, 4.2.2.3 and 4.2.2.4. ,

Criticality Treatment of Interaction between Suberitical

4.2.10 Units 4.2.10.1 Equipment and facilities may be considered to be

nuclearly isolated if they are separated by either of

the following:

7c A one foot slab of water or by the distance which is equivalent in isolation ability to a one foot slab of water.

The larger of 12 feet or the greatest distance across an orthographic projection of the largest of the fissile accumulations on a plane perpendicular

L to the line joining their centers.

The criticality effects of the exchange of neutrons

4.2.10.2 between individual suberitical units which are not

isolated may be treated by either of the following

techniques:

Techniques which produce a calculated

multiplication factor of the entire system (e.g.,

Monte Carlo) may be used. When this is done, the

analysis must comply with the requirements of

Sections 4.1.1, 4.2.2.3 and 4.2.2.4.

l

=I LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.21

Techniques which do not produce a calculated

multiplication factor for the entire system but

instead compare the system to accepted criteria *

(e.g., Solid Angle technique) may be used. When

this is done, the analysis must comply with the

requirements of Sections 4.1.1 and 4.2.2.4.

Requirements for Engineered Controls

4.2.11 Engineered controls detect an undesired situation and implement corrective action without requiring human

Engineered controls ntust be:

intervention.

Sufficiently dependable and so used that the

probability of failure is minimized.

Capable of performing the criticality safety

, purpose for which they are specified.

Maintained and/or calibrated on a schedule suitable

for the specific device and the specific

application.

Verified as being properly installed prior to first

use with fissile material.

Modified only with documented, prior approval of

the criticality safety function.

Supported by procedures and/or devices which

provide: continued control if operation is to be

allowed to continue after a control has failed.

l l

t l ,

l=5' SNM-1097 10/23/84 LICENSE DATE PAGE DOCKET 70-1113 REVISION ~ 6 I-4.22

1 So designed that when sampling is part of a control, the sampling is performed at a frequency consistent with the rate of variation of the

parameter and with the implementation of control

action.

4.2.12 Requirements for Procedural Controls A procedural control requires human intervention in

detecting an undesired condition and/or implementing corrective action. Procedural controls must be:

Sufficiently depen'dable and so used that the

probability of failure is minimized.

Capable of performing the criticality safety

purpose for which they are specified. *

(

Implemented by formal written procedures.

Shown to be complied with by formal written records.

Modified only with documented, prior approval of the criticality safety function.

i Supported by procedures and/or devices which

provide continued control.if operation is to be allowed to continue after a control has failed.

So designed that when sampling is part of a l

' control, the sampling is performed at a frequency

i

-s LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION. 6 I-4.23

's ,

consistent with the rate of variation of the

parameter and with the implementation of control

action.

4.2.13 Criteria for Fire Protection in Areas Containing Fissile

Material 4.2.13.1 Fire protection shall be provided for equipment,

processes, and facilities containing fissile material and shall be selected on the basis of minimum impact on area nuclear criticality safety.

4.2.13.2 The use of water for fire protection in moderation

control areas shall be minimized and controlled.

4.2.13.3 Fire protection instructions covering the manufacturing

facility are issued which communicate necessary or

, permissible methods or techniques to be used.

4.2.14 Incineration of Nuclear Waste

The incinerator is authorized to operate with an estimated uranium enrichment of not less than 2.75% in U235 for batch control. Waste boxes with assigned enrichments greater than 2.75% in U235 can be incinerated with proper demonstration of safety provided the boxes are generated from process areas where the maximum nominal enrichment handled is not more than 1

4.0%. No waste boxes generated from a process which operates with a maximum enrichment greater than 4.0%

nominal will be incinerated unless physical measurements l

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.24

'o of the U235 content are made or the highest enrichment

.i in the box is assigned to the batch.

(

t.

LICENSE SNM-1097 DATE 10/23/84 PAGE DOCKET 70-1113 REVISION 6 I-4.25

~

f $ - //0 GENER AL $ ELECTRIC W1LMINGTON MANUFACTURitt3 DEPARTMENT GENERAL ELECTRIC COMMNY PO. BOX 780 + WILMINGTON, NORTH CAROUNA 28402 December 10, 1984 co V cy Director '9 g

Office of Nuclear Material Safety & Safeguards RECENED U. S. Nuclear Regulatory Commission g7 I Washington, D. C .- 20555 DEC11 $ $ -li 1

Attention: Mr. W. T. Crow, Section Leader V "$,f[s Uranium Process Licensing Section ud secto

M/C 396-SS Cv .-

to Gentlemen:

References:

(1) NRC License SNM-1097, Docket #70-1113 (2) Letter, CM Vaughan to WT Crow, 6/1/84 .

(3) Letter, GH Bidinger to CM Vaughan, 9/11/84 '

Subject:

REQUEST FOR WITHDRAWAL OF PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE On June 1, 1984, General Electric Company submitted an application to the NRC requesting authorization to operate equipment and processes known as UPMP (Uranium Process Management Project). -

That submittal included two versions of the application - one with .

proprietary information included and one for release to the NRC Public Document Room with proprietary information deleted.

During a recent review of the 6/1/84 application, it was discovered that several pages in the nonproprietary submittal inadvertently contained proprietary information. Therefore, General Electric requests that this information be withdrawn from the Public Document Room (s) and returned to GE.

In order to facilitate withdrawal and replacement, a complete copy of the 6/1/84 non-proprietary submittal is attached which incorporates revised.pages. All information deleted from these revised pages was designated as proprietary by the affidavit accompanying the submittal (J. A. Long, 6/4/84) and confirmad by the NRC (Reference 3). Attachment I contains a list of the effected pages.

General b=

Electric Company requests that this letter and Attachment 1 =ithheld from public disclosure until such time as the information is re:aoved from the PDR(s).

l

@M Sy&R s

%n$,,y e i a un 9 y5.36

~

e M w tN coi. .n f 5 - I 5 f d_ -

GENERAL $ ELECTRIC Director - ONMSS December 10, 1984 Page 2 General Electric Company personnel would be pleased to discuss this matter with you and members of your staff as you may deem necessary.

Very truly yours, GENERAL ELECTRIC COMPANY t4b hk gq -

Charles M. Vaug an, Acting Manager Regulatory Compliance ,

M/C J26 CMV:bsd I

NSD/SGD-L Attachment b

6

- - . - .- , , - - - - ,e,-- - -., . - , , ,,,_r -- ,-v - ,-

GENER AL h ELECTitlC Director - ONMSS December 10, 1984 ATTACHMENT 1 The following pages in the 6/1/84 non-proprietary submittal for the UPMP project have been revised to delete proprietary information:

2.1-13 ;

2.2-15 2.2-17 2.3-19 '

2.4-27 2.4-29 2.5-12 4-27 4-29 4-30 4-47 4-48 CM Vaughan

bsd

~

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

Si cocxzr 30. 70- /N 3 CONTROL NO.

M Ob _ _,

DATE OF DOC. /8[/4/8h DATE RCVD. / r

!/!

/ /

FCUF V f _ PDR_

4 FCAF LPDR UM- I&E REF. [

WiUR SAFEGUARDS F..T C OTHER f,pCRIPTION:

Alds& W &

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M L/ bt2 S nslb At!)

%xkw 2 /O!cf '

INITIAL

/ /

~

{ G E N-E R A L ELECTRIC COMPANY i

AFFIDAVIT i

I, James A. Long, being duly sworn, depose and state as follows: !

I

'; 1. I am General Manager, Wilmington Manufacturing Department, General Electric Company, and have been delegated the function

of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withhold,ing. ., ,,

2.- The information to be withheld is identified portions of the !

Uranium Process Management Project (UPMP) licensing {

demonstration document dated June 1, 1984, supporting [

Application Amendment N-18/S-32, NRC License SNM-1097, Docket '

470-1113. This licensing action was first noticed with the Nuclear Regulatory Commission 14 December 1982.

3. In designating material'as proprietary, General Electric !

utilizes the definition' of proprietary information and trade ; !

secrets set forth in the American Law Institute's Restatement' !

Of Torts, Section 757. This definition provides: i A trade secret may consist of any formula, pattern, i device or compilation of information which is used in j

one's business and which gives him an. opportunity to obtain an advantage over competitors who do not know or t use it.... A substantial element of secrecy must exist, [

, so that, except by the use of improper means, there would [

_ __ . _ _ be dif ficulty in acquiring information. . . . Some factors ;

to be considered in determining whether given information is one's trade secret are: (1) the extent to which the information is known outside of his business; (2) the t extent to which it is known by employees and others involved in his business; (3) the extent of measures [

i taken by him to guard the secrecy of the information; (4) i the value of the information to him and to his i competitors; (5) the amount of effort or money expended [

by him in developing the information; (6) the ease or f difficulty with which the information could be properly I acquired or duplicated by others." !

4. Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method or apparatus I where prevention of its use. by General Electric's I competitors without license from General Electric constitutes a competitive economic advantage over other companies; ,

l

.s j Y

1- I I

i

i t i

-- -. , - - . , . , , , - , , . , , - , - - - - - , , , , . ,..,,--.---,-.-._--,--.,.--n.-,--n.- - , - - - - . - - - - _ - - - - - _ _ _ . - -

l i

b. Information consisting of supporting data and analyses, l including test data, relative to a process, method or i apparatus, the application of which provide a competitive economic advantage, e.g. , by optimization or improved marketability;

c. Information which if used by a competitor, would reduce his expenditure of, resources or improve his competitive i position in the design, manufacture, shipment, l installation, assurance of quality or licensing of a <

Reference:

~ SNM-1097

June 1, 1984 TABLE OF CONTENTS Section Description Page 2.3.2 UPMP Nitrate Waste Treatment 2.3-3 2.3.3 Criticality Safety considerations 2.3-12 2.4 SCRAP PROCESSING 2.4-1 2.4.1 Current Scrap Processing Operation 2.4-1 2.4.2 UPMP Scrap Processing 2.4-1 2.4.3 Criticality Safety considerations 2.4-25 2.5 URANYL NITRATE CONVERSION 2.5-1 ;

2.5.1 Current Conversion Process 2.5-1 2.5.2 UPMP Conversion Process 2.5-4 2.5.3 Criticality Safety Considerations 2.5-6 ,1 2.6 _

PROCESS CONTROL SYSTEM 2.6-1 2.6.1 System Power 2.6-1 2.6.2 control System Features 2.6-1 2.6.3 Systen Operation 2.6-3 2.6.4 Operator Consoles 2.6-4 2.6.5 Hard Copy Printout 2.6-4 CHAPTER 3 - SUPPORT SERVICES & EQUIPMENT 3.1 SOLVENT TREATMENT 3-1 3.1.1 Geometry Control of'the Solvent Treatment 3-2 System i

f I

l l

Reference:

SNM-1097 l

June 1, 1984 i TABLE OF CONTENTS Section Description Page t

3.2 FILTER CLEANING STATION 3-3 i 3.2.1 Geometry Control of the Filter Cleaning 3-5 Station (

3.3 NO, ABSORPTION 3-5 i 3.3.1 Criticality Considerations 3-8 !

l 3.4 AQUEOUS AND SOLVENT MAKE-UP (ASMU) 3-10 f

i 3.5 SOLID WASTE HANDLING 3-10 ;

9 3.6 PROCESS LABORATORY 3-12 ,

I CHAPTER 4 - CRITICALITY SAFETY ANALYSES t POR EQUIPMENT & OPERATIONS !

4' .1' GE-WMD CRITICALITY SAFETY POLICY & METHODS 4-1 j 4.1.1 Double Contingency Policy 4-1 4.1.2 Geometry Control 4-1 4.1.3 Fixed Neutron Absorbers 4-2 4.1.4 Administrative Controls 4-3 l 4.1.5 Definition of Critically Safe 4-3 j 4.1.6 Analytical Methods Used in Criticality 4-4 Safety Analyses 4.1.7 00 2 & Water Mixtures 4-4 .

4.1.8 The Reactivity Formula 4-7 l

t 111 -

Reference:

SNM-1097 '

t i

June 1, 1984 i

TABLE OF CONTENTS Section Description Page 4.1.9 Infinite Neutron Multiplication Factor 4-7 (Kinf) Considerations 4.1.10 Equipment Interaction 4-15 4.2 UPMP SAFE GEOMETRIES & SAFE BATCH LIMITS 4-17 4.3 CRITICALITY SAFETY OF UPMP CYLINDRICAL

PROCESSING VESSELS 4-17 4.4 UPMP FIXED NEUTRON ABSORBER PANELS 4-22

3 4.5 CRITICALITY SAFETY OF UPMP PROCESSING VESSELS 4-27 4.5.1 Vessel Descriptions 4-27

, 4.5.2 Neutron Absorber Panels 4-29 ,

4.5.3 Analytical Methods 4-29 4

--4.5.4- Results 4-39 i

4.6 CRITICALITY SAFETY OF UPMP DISSOLVER

& LEACHING VESSELS 4-46 4.6.1 Dissolvers 4-46 ,

4.6.2 Leachers 4-58 4.7 CRITICALITY SAFETY OF UPMP SOLVENT !

EXTRACTION COLUMNS 4-62 4.7.1 Solvent Extraction Column Geometry 4-62

, 4.7.2 Analytical Methods 4-62 4.7.3 Results 4-65 l - iv - ,

Reference:

SNM-1097

June 1, 1984 TABLE OF CONTENTS Section Description Page i

4.8 CRITICALITY SAFETY OF UPMP THREE &

FIVE GALLON PAILS 4-65 4.8.1 Container Geometries 4-65 4.8.2 Analytical Methods 4-65 4.8.3 Results 4-67 4.9 CRITICALITY SAFETY OF UPMP FURNACE BOATS 4-67 l 4.10 CRITICALITY SAFETY OF UPMP ROLL CRUSHER 4-67 4.11 CRITICALITY SAFETY OF PROCESS SUMPS &

FLOOR BASINS 4-72 4.11.1 Geometries 4-72 4.11.2 Analytical Methods 4-72 4.11.3 Results 4-74 4.12 CRITICALITY SAFETY OF UPMP EQUIPMENT INTERACTION 4-77 4.12.1 Monte Carlo Code Applications 4-77 '

4.12.2 Solid Angle Code Applications 4-77 4.12.3 Isolation by Distance 4-78 !

4.12.4 Isolation of Process & Dissolver 4-78 Vessels Criticality Safety of UPMP for Process

~

4.12.5 Piping Interactions 4-80 i L

f !

-v-

! Refercnce: SNM-1097 i

June 1, 1984 TABLE OF CONTENTS Section Description Page CHAPTER 5 - RADIOLOGICAL SAFETY CONSIDERATIONS 5.1 BUILDING VENTILATION SYSTEMS 5-1

5.1.1 Recircu'lation Systems 5-1 t

5.1.2 Equipment Enclosures 5-2 5.1.3 Absolute Filter Systems '

5-2 !

5.1.4 Ductwork 5-3 !

l 5.2 EQUIPMENT-OPERATION 5-3 ;

f f

5.2.1 Facility Exhaust System 5-3 5.2.2 Solvent Extraction & Dissolution Room 5-6 Make-up and Recirculation System 1

5.2.3 Process Area Recirculation System 5-6 3 l

1 5.2.4 Process Area Make-up Air System 5-7 '

5.2.5 ASMU Operating Area (Non-controlled) 5-8 l [

5.2.6 ASMU Chemical Mix Area (Non-controlled) 5-8 5.2.7 Control Room and Office (Non-controlled) 5-8 +

r 5.2.8 HEPA Filter Banks 5-8 i

l 5.3 PERSONNEL EXPOSURE CONTROLS 5-9 ;

l I

5.3.1 Measurement of Air Concentrations 5-9 ;

5.3.2 Contamination controls 5-10 5.3.3 Criticality Detection & Evacuation Alarm 5-10 i System !

\

! 5.3.4 Operating Instructions to workers 5-11 l

l l - vi - !

)

Reference:

SNM-1097 I

~

w

?%

June 1,~1984 -

s ,

TABLE OF CONTENTS s 1 Section- Description _ Page CHAPTER 6 - ENVIRONMENTAL CONSIDERATIONS 6'.1 OVERVIEW -

6-1 6.2 EFFLUENT QUALITY -

, 6-3 6.2.1 Treated ProEess Liquid Effluents '-

6-3 6.3 AIRBORNE _ EMISSIONS 6-8 4

6.4 SOLID WASTE -

~

6-9. ;

Sludges from. Fluoride', Rad Waste &

6.4.1 6-9 Nitrate Wastes

.q 6.4.2 Other Solid Wastes & Sludges 6-10

\

6.5 . REDUCED TRANSPORTATION RISK ' '

6-11 s s. s

^6.6~ ~ CLOSURE AND DECOMMISSIONING b 6-11

\ s\

t APPENDIX ,

GENERAL ELECTRIC COMPANY DRAWINGS

<- , a , .

s AR-001, Revision 6 % Equipment Arrangment Plan, 1st Floor

<. North - UPMP Area l

AR-002, Revision 6 Equipment Arrangement Plan, 1st filoor

. South - UPMP Area s

AR-003, Revision 5 Equipment Arrangement Plan, Mezzanirse UPMP Areas -

s

.s AR-004, Revision 5 Equipment Arrangement Plan, 2nd' Floor l North - UPUP Area

\ - , ,

x l

. 5 '.-

. t-

% 4 l "

's - -

s o g-vii -

Reference:

SNM-10974 s

- i , s s

.. . s s, \,;

T g(,

s , . _ . __

June 1, 1984 i TABLE OF CONTENTS Section Description Page AR-005, Revision 5 Equipment Arrangement Plan, 2nd Floor South - UPMP Area AR-005, Revision 5 Equipment Arrangement Plan, Solvent Extraction Cell Area AR-007, Revision 5 Equipment Arrangement Section, UPMP Area AR-008, Revision 3 Equipment Arrangement Section, UPMP Area AR-015, Revision 5 Equipment Arrangement, Chemical Mix &

Main Process Scrubber Areas

, i, I.

- viii -

Reference:

SNM-1097

qune 1, 1984 CHAPTER

1.0 INTRODUCTION

1 This document provides a process description and demonstration of safety for the Uranium Process Management Project (UPMP) facility expansion at the General Electric Company, Wilmington Manufacturing Department, Wilmington, North Carolina. This new facility will be, operated under Special Nuclear Material License SNM-1097 and will become an integral part of nuclear fuel production at the Wilmington site.

1.1 PURPOSE The purpose of UPMP is to establish the recovery and recycle of uranium from waste ' streams and on-site uranium scrap recovery as normal processes within the current fuel fabrication operation. -

i, 1.2 GENERAL PROJECT DESCRIPTION 1.2.1 Project Scope The UPMP includes six basic process and control segments which have been synergistically integrated to recover uranium from off-line material streams and to convert the recovered uranium to uranium ~ dioxide powder for use in standard nuclear fuel products. The project also includes chemical and other related support systems and building modifications.

The basic UPMP segments include:

1.2.1.1 Fluoride Waste Treatment ,

A treatment process to remove low concentrations of uranium from the ammonium Etuoride liquid waste stream which results from the chemical conversion of uranium hexafluoride to uranium dioxide.

1.2.1.2 Rad Waste Treatment A lime treatment and solids separation process to remove low concentrations of uranium from a waste stream ,

composed of laundry water, decontamination wash water, floor mop water, incinerator scrubber water and laboratory drain discharge,s.

l .

Page 1-1 -

Reference:

SNM-1097

i

~

i June 1, 1984 1.2.1.3 Nitrate Waste Treatment f i

A two part process consisting of (1) a primary lime treatment and uranium solids separation system for the ;

ammonium nitrate liquid waste stream generated in the ;

conversion to uranium dioxide from uranyl nitrate and ;

3, (2) a secondary lime treatment and solids separation !

-i system to handle the nitrate waste treatment effluent t and the. waste stream from the UPMP solvent extraction i system. The combined nitrate liquid wastes will be t neutralized, treated for removal of metals and ;

ultimately released to the existing site nitrate lagoon. j Contaminated solids will be packaged for shipment to , i licensed offsite burial facilities. !

e 1.2.1.4 Scrap Processing A uranium scrap pre-treatment, (primarily oxidation), .

nitric acid dissolution and solvent extraction process' '. !

to recover uranium from scrap materials and sludges. ,e' l This process will produce a pure uranyl nitrate product I suitable for conversion to nuclear reactor grade uranium i dioxide. j 1.2.1.5 Uranyl Nitrate Conversion ; i l

A uranyl nitrate conversion (to uranium dioxide) process. !

This process will replace the existing batch hydrogen !

peroxide precipitation process.

1.2.1.6 Process Control Systems )

The UPMP process control system il a distributed system with dual consoles. Each console will monitor one-half !

of the UPMP process while providing backup for the other !

half. Power is provided via an interruptible power i supply with battery backup. j i

1.2.2 Facility Description !

i As shown in Figure 1-1, the UPMP facility is located in I the northwest c'orner of the existing high bay FMO-X l building and in three new contiguous building additions i i

l I

i i

Page 1-2 CD

Reference:

SNM-1097

's

! . _ ..__-_m,

June 1, 1984 FIGURE 1-1 UPMP FACILITY er -

sw/

-e- b f , _ . _,, .. . -

n m /// :) 5, Q.,/M / / '

g

/' s s*.*.*.e j ee.m ese m,,,,,, , , , , , , , , , , , , , ,

- =, .

/ -

,--" 1.__. . .

_Q O = i. g := .

_e &n ,

O x4 -

] E "me*.** .

aos

~, -

g meee wn GeOUNOFLOOR E3 :: .*.::',: ,

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l Page.1-3 l

SNM-1097

Reference:

l l

June 1, 1984 1

to the north and west. The total UPMP process and support floor area is approximately 29,200 square feet.

Modifications to the existing building consist of the l addition of a second floor and a mezzanine level for the

! control room, computer room, and process laboratory.

l The three new buildings are a three story solvent i extraction building, a solvent and chemical makeup building, and a single level structure for weather protect. ion of the electrical substation and uninterruptible power supply system (see Figures 1-2 through 1-4).

1.2.3 Integration with Existing Manufacturing Facilities ,

Figure 1-5 schematically presents the current fluoride, nitrate and rad waste systems. Figure 1-6 indicates how the new UPMP facility will fit into the current waste handling systems. As illustrated in this Figure, the ,

new UPMP facility will provide integral pre-treatment '.

and in-line waste treatment systems to reduce the waste stream uranium content and to increase subsequent uranium recovery.

Selected utilities and services will be supplied from or connected to existing FMO-X systems. These include electrical power, fire detection and sprinkler systems, }

steam, chilled water, potable water, compressed air, instrument air, process gases, storm sewers and process sewers. In addition, UPMP nuclear safety systems (e.g.,

criticality alarms and air sampling) will be connected to existing facilities.

1.2.4 New Processes and Technological Features The UPMP is a extensive combination of industrial processes and technological features. Several are new in WMD operations and are of specific note in this license amendment application; as indicated below, several are also considered proprietary to GE.

o A system for fluoride waste treatment (trade secret) o A solvent extraction system for uranium recovery and purification I o An uranyl nitrate conversion process (patent to be I applied for)

Page 1-4

Reference:

SNM-1097

June 1, 1984 FIGURE 1-2 UPMP EQUIPMENT LAYOUT - GROUND FLOOR

- 8OLVENT SXTRACTION COLUMNS NITRATE WASTE COLLECTION TANK

(

EXYAACTION d

SOLVENT EXTRACTION FEED TANut t >A tL swA 8T'O LINE SLURRY TANK 8 fSOX HANDLINQ 3

' 4* <' OO :arafiO5 C

.Nr.,,5 WA.T . ,~ ' Q: o clasOLVLNG

"* **" **"" ])I i.

[C ,,,,, / [ ETAE " " "

RAD WASTE TANK y =* C enOCesslNo , /

i PRODUCT TANKS o 'O O O Q38 .1[;"' ~ -

f ,i c ~

"^' *^sTE TANS -

O j O

j QQ

~"'"'""'"""^*'

0 : r[o ..AOf 3

FLUORIDE WASTE TANK :$ v

-O *' :~ '="aYT '*

E 6 ^"""*"^*^""*

SHOP at

ELECTRICAL SUBSTATION AR'EA s O i .:. O _.~

- A w,$ ,

ea ~

RAD WASTE i .

M s. 5S C ' ' '

RECEIVER TANK 4 ' C ,

FLUORIDE WASTE UNH SLAS STORAGE TANKS TANKS I

l Page 1-5 l

Reference:

SNM-1097 I l

1

June 1, 1984 FIGURE 1-3 UPMP EQUIPMENT LAYOUT - MEZZANINE n

cowuran uo Roou s -

-- _j_ _

orr:Cs _

_3 C

"" - 3 PROCESS l CONTROL ROOM LA8 J l- I h ,

/

~..

~

.. s,

. . ~ .

g Page 1-6

Reference:

- SNM-1097

1 i

l June 1, 1984 i

FIGURE 1-4 UPMP EQUIPMENT LAYOUT - SECOND FLOOR SOLVENT EXTRACTION COLUMNS

/

o'ci SOLVENT d

E X T R ACTION GX/DtSSOLVER ROOM RECIRC. FAN SLOWER g ] M jl 9C CAN LIFT MOTOR CONTROL CENTERS ( d-

{ft_J ~D S LTER SX/DISSOLVER SCRUSSER

/ I OAD PROCESS SCRUSSER

-V I OAD RECIRC ASS FILTER OAD RECIRC. SLOWER NITRATE & RAD WASTE RECtRC. PUMPS Ia 1

[O HEPA FILTERS

'00000'

[

CHILLER -

i l

l I

Page 1-7

Reference:

SNM-1097 !

l

June 1, 1984 FIGURE 1-5 EXISTING FLUORIDE, NITRATE & RAD WASTE SYSTEMS PEATING ,

AIRCRAFT ENGINE WASTE

AND EQUIPMENT PRECIPITATION ETCH' TREATMENT MANUFACTURING WASTE' BUILDING SLUDGE SPENT OFFSITE NITRIC DISPOSAL e

SPENT ETCH 9r ,

OFFSITE FUEL DISPOSAL COMPONENTS OR USE MANUFACTURING OFF BUILDING O SITE SPENT SODIUM USAGE ;

HYDROXIDE ,

NITR ATE FINAL DISCHARGE a FUEL WASTE 9 SAMPLE POINT %

MANUFACTURING PRECIPITATION BUILDING TREATMENT gY j JiLUDGEl FOR NITRATE LAGOONS lEWORK 3 m FLUORIDE I

" WASTE { I

- ~~'

SLUDGE 4 AMMONIA p f sp SEPARATIOP RECOV 5' m

SLUDGElFOR REWORK 1r f 7 y' FLUORIDE m

RECOVERED LAGOONS J E

! AMMONIA fj $

RAD W A STE 7 AERATION LAGOONS e

\ ADJUST FINAL PROCESS LAGODNS l

FINAL DISCHARGE SAMPLE POINT -

x -;'.

DAM

d TO RIVE l

Page 1-8

Reference:

SNM-1097

l o

June 1, 1984 FIGURE 16 UPMP FLUORIDE, NITRATE & hAD WASTE SYSTEMS PEATING m WASTE AIRCRAFT ENGINE PRECIPITATION AND EQUIPMENT ETCH ' TREATMENT M ANUFACTURING W A S T E' BUILDING vsLUDGE SPENT OFFSITE NITRIC DISPOSAL SPENT ETCH OFFSITE FUEL DISPOSAL COMPONENTS OR USE OFF MANUFACTURING g

?

W SITE BUILDlHG i ) ' USAGE SPENT SODIUM HYOROXIDE .

FINAL DISCHARGE ag ',

REACTION

SAMPLE POINT %

$ SOLIDS REMOVAL MANUFACTURING BUILDING g jg j

, NITRATE LAGOONS SOLIDS ab 'E TO BURIAL

(>U rm

- FLUORIDE EACTIOb RECYChi. v w337g i z AND LUORIDE AM M O NI A 9 N SOLIDS {REATMEN gT RECOV 5' EMOVALa , '

"/b Qw

C e

SCRAP T iP 4 FLUORIDE DROCESS 2 RECOVERED LAGOONS J E W '

AMMONIA y $

AD; W ASTE AERATION LAGOONS e

\ b AbJ1JST FIN AL PROCESS LAGODNS, T'NAL DISCHARGE hMPLE POINT KEY:

UPMP DAM PROCESS .

4' STEPS TO RIVE l

l 1

I l Page 1-9 l l

Reference:

SNM-1097 l l

l -- - - -.

June 1, 1984 o A scrap dissolving process for low grade sludges and incinerator ash _

t o Lime precipitation and solids removal in rad waste 4 treatment (details of the process are a trade secret). ,,

. o Lime precipitation and solids removal in nitrate waste treatment (details of'the process are a trade i secret) o Geometrically . safe vessels and scrap dissolvers utilizing neutron absorber ,

panels (trade secret) o Automated mechanical filters for solids removal l

o -A distributed process control system ,

o' A dedicated process laboratory which includes t t

for process analytical measurements (patentable applications) o Bubble cap tray design absorption column for removal i of oxides of nitrogen from dissolver offgas 1 o Tandem instrumentation for in-line uranium and acid measurement The following chapters in this document present a detailed description of the processes and features listed above.and highlight not,only the process designs but also the criticality, radiological and environmental safety considerations.

Included as an appendix -to this document is a package of facility and equipment drawings. These drawings are particularly useful in detailed reviews of information

in Chapters 2-5 of the demonstration since the equipment numbers defined in the text are consistent with the numbering on the drawings.

1.3 PROJECT BENEFITS UPMP processes have been extensively tested and, as a result of these development efforts, success in achieving the predicted project benefits is assured.

Operation of UPMP will have significant positive environmental impacts and reduced safety risks.

Page 1-10

Reference:

SNM-1097

June 1, 1984 1.3.1 Improved Radiological Environmental Protection The quantity of uranium in radioactive liquids (fluoride, rad waste, and nitrates) discharged from the manufacturing operation will be reduced approximately

4. t

! The uranium concentration in calcium fluoride sludge will be. reduced sufficiently to consider disposal of the sludge under Options 1 and 2 of SECY 81-576 or for chemical use in non-nuclear operations.

The quantity of nuclear material to be disposed of for

,' site decommissioning will be reduced.

1.3.2 Improved Non-radiological Environmental Protection 4

UPMP reduces generation of calcium fluoride sludge from rad waste by 4. Therefore, not only is the uranium '

t concentration in the sludge reduced, but also the quantity of this by-product material is minimized.

The facility design extensively treats process effluents prior to their discharge to the atmosphere. Although the atmospheric discharge from the WMD site will be increased insignificantly as a result of the additional material being processed, the overall discharge to the atmosphere, when the current offsite reprocessing

- -- ~_. -operations are considered, will be reduced. The net result will be a favorable environmental impact for the Fuel Cycle.

The UPMP process designs have focused on resource conservation. Solutions in solvent extraction are

processed so as to permit recycling through the operations thus reducing the quantity of waste chemicals and the need for their disposal.

1.3.3 Reduction of Risks Currently scrap material is shipped over-the-road for off-site reprocessing; UNH solutions are returned to Wilmington. Since UPMP implements in-house scrap i

reprocessing, external shipments of these nuclear materials will no longer be needed. This eliminates the

, risk of potential nuclear releases to the environment during accidents on the highway.

Page 1-11

Reference:

SNM-1097 ,

June 1, 1984 A new conversion process eliminates the use of hydrogen peroxide, thus avoiding a potentially hazardous industrial safety condition.

1.3.4 Contribution to Site Decommissioning Complete removal from the site of all contaminated sludges is a commitment in the Closure and Decommi.ssioning Plan approved by the NRC in December, 1981. Approximately cubic feet of uranium- t bearing calcium fluoride sludge have been accumulated and stored on site since the plant started up. t t

t t

t Thus UPMP is, in essence, the first step in t completing the decommissioning activity. . ;

I i

Page 1-12 l

Reference:

SNM-1097 .

June 1, 1984 CHAPTER 2.0 PROCESS DESCRIPTION 2.1 FLU' RIDE WASTE TREATMENT SYSTEM 2.1.1 Current Operation i

In the conversion of uranium hexafluoride to uranium dioxide, an ammonium fluoride waste stream is generated which contains a very low concentration of uranium.

The fluoride waste liquids resulting from the Fuel Manufacturing Operation (FMO) conversion processes are currently accumulated in quarantine tanks and are circulated through a filtration system to take out the uranium solids. (See Figure 2.1-1.) The clear liquid from the system is accumulated in the quarantine tanks

, and is sampled for uranium concentration. If the . ;

uranium concentration does not exceed the release limits for liquid transferred, it is released from the quarantine tanks to the accumulation tank located outside of the building. The sludge is accumulated in five gallon pails, transferred to the REDCAP drying furnace in FMO and is subsequently recycled through an existing batch uranyl nitrate conversion process (UPS).

The fluoride waste stream is batch pumped from the V-106 tank to a 100,000 gallon settling tank, (V-108), at the process Waste Treatment Facility. Any uranium solids that settle to the bottom of the V-108 tank are removed by centrifuging. This wet sludge is packaged in five gallon plastic pails and is transported to the REDCAP furnace area to be dried. The clear supernate from V-108 is decanted and is treated with lime to precipitate calcium fluoride and uranium. The stream is steam stripped of ammonia and pumped into a lined lagoon where the calcium fluoride and uranium solids settle to the bottom. The supernate from the lagoon, which is nearly free of fluoride, ammonia ~and uranium, is discharged to the final process lagoons.

Page 2.1-1 l

Reference:

SNM-1097

June 1, 1984 FIGURE 2.1-1 FLUORIDE WASTE FLOW - PRESENT PROCESS u

W QUARANTINE Slab Tanks SLUDGE &

1/

FILTER V .

QUARANTINE Slah Tanks 4

q ppm U t TANK V106 ,

I g

ppm U t SLUDGE b TANK V108 Ppm U t 9

LIME : AMMONIA 4 AQUA NH 3 RECOVERY d

LAGOON Y

y <1 ppm U SLUDGE ppm U Clear Effluent t to Final Process t Lagoons l

l l

l Page 2.1-2 l

Reference:

SNM-1097 i I

m . . . ._ - . . _ . _ _ ._ __ _ _. ___ . _ _ _ . _ _ . _ . . _ .

June 1, 1984

A side stream of fluoride waste, is diverted from tank

- V-106 to the existing GECO rundown tanks. These two tanks contain about 2500 gallons of waste water. A very high volume is recirculated from these tanks through

, high capacity ejectors then back into the tanks. The purpose of this system is to develop a vacuum for the GECO conversion process and to scrub the process gasses.

A liquid flow equal to the input flow is returned to

! tank V-106. The operation of the rundown tanks is

controlled by the GECO process control system.

, 2.1.2 UPMP Fluoride Waste Treatment System

. . To further reclaim uranium from the fluoride stream, (mostly soluble or colloidal), and to minimize the uranium concentration in the calcium fluoride sludge, a system will be installed between existing tanks V-106 i and V-108 for processing the fluoride waste. (See Figure 2.1-2.) This new process is described as , ;

j follows: .

i 2.1.2.1 Surge Tank - Existing Tank V-106 The clarified fluoride waste liquid from the FMO

quarantine tanks.will be accumulated in the surge tank V-106 which serves as the feed tank for the fluoride waste treatment system. The process line from the i quarantine tanks to V-106 will not be changed from the existing system and transfers to tank V-106 will be based on the existing release limit controls. One i difference, however, will be that the density of the i liquid in V-106 will be monitored by the UPMP process

control system ' and concentrations of greater t

than grams of uranium per liter will result in a 1 cut off of the feed from the FMO quarantine' tanks.

1

. The waste liquor in the surge tank will contain !

L compounds such as ammonium fluoride, ammonium hydroxide, l' insoluble and colloidal uranium t I

compounds and trace quantities of nitrates and i Before this solution can be processed t ,

through the fluoride waste treatment system, the insoluble'and colloidal uranium compounds must be ,

dissolved. The waste liquor in V-106 will be kept in I F i h

4

I Page 2.1-3 !

l Reference SNM-1097 ;

l

June 1, 1984 FIGURE 2.1-2 FLUORIDE WASTE FLOW - UPMP PROCESS so 4 QUARANTINE Slab tanks SLUDGE d V

FILTER v

QUARANTINE Slab tanks y

ppm U . t TANK SOLUTION --D V-106 y

ppm U t 1

SOLUTION RECOVERY --a PRODUCT -> SCRAP J MAKEUP ---e SYSTEM g ( ppm U) RECOVERY t y

ppm U t TANK I v-108 Ppm U t :

3, LIME ---> AMMONIA ---> AQUA NH 3 RECOVERY ;

t ppm U V '

LAGOON --

q q ppm U Key: t SLUDGE CLEAR EFFLUENT UPMP (Dry Basis) TO CHEMICAL LAGOON l i

t t

{

Page 2.1-4 t

Reference:

SNM-1097

June 1, 1984 constant circulation by a dedicated high volume pump.

This will prevent the insoluble uranium compounds from settling. This circulating stream flows through a venturi type gas scrubber on the top of V-106 which is used to mix the tank contents, f

A separate pumping system which utilizes parallel pumps, will be used to transfer the fluoride waste solution from the surge tank to the system, through the filters and, ultimately, to tank V-108 at the Waste Treatment Facility.

At the design flow rate, the parallel pump system relied on for transfer will not have sufficient head to overcome the total pressure drop through the system. To assure adequate, stable operation, a booster pump has been added to the feed system. The booster pump will increase the pressure of the feed to compensate for the pressure drop across the system. Uranium is either ;

collected or caught in the filter. The effluent filter '

also prevents any solids from passing through to tank V-108.

2.1.2.2 Recovery System The process will be a t system, operating with a t The operation t and of the will be t controlled by the UPMP process control system. One complete cycle includes the following sequences: (See Figure 2.1-3.)

2.1.2.2.1 t 2.1.2.2.1.1 t t

t t

t l t i

. t

, t l

l l

i l Page 2.1-5 Reference SNM-1097

June 1, 1984 P

FIGURE 2.1-3 FLUORIDE WASTE TREATMENT PROCESS FLOW L

t t

t T

t t

. ; t t

4 t t

i t

t

't i

t t

! 1

! t t

t t

t t

1 i

i Page 2.1-6 Reference SNM -1097

~

\

a

~

June 1, 1984 t

t

1 t

t t

- t 2.1.2.2.1.2- t j t t

t t

t t

t 2.1.2.2.1.3 t j t t

2.1.2.2.2 .

t t

[ t

t 4

t t

i t i t t

t

] 2.1.2.2.2.1 t i

! t i t t

t j t

't i -

t j t t

2.1.2.2.2.2 t t

t

- . t t

t i

i l

l Page 2.1-7

Reference:

SNM-1097

June 1, 1984 t

t t

t 2.1.2.2.3 t t

t t

t 1

t t

. ; t 9

2.1.2.3 Product Adjustment

.The product, as it is being received into the product vessel, will be mixed with nitric acid at a controlled rate. The rate of acid addition will be automatically ..

controlled to maintain the mixture at the proper pH. l The nitric acid will be supplied from a nitric acid

. break tank which will be a critically safe pipe tank that serves to isolate the acid from the uranium bearing process.

The product, after it is acidified, is circulated in the vessels to mix the solution thoroughly and will be

-sampled and chemically analyzed.

f The aluminum nitrate will be supplied from a critically safe' aluminum nitrate break tank. This tank will isolate the supply system from the uranium bearing i process.

l The product, which contains about kgs of uranium, t i will be either'used in the scrap processing dissolvers l

i l

l Page 2.1-8 References SNM-1097-

4 June 1, 1984 l

and leachers or, if the dissolver/ leachers are not 4

operating, will be processed through the nitrate waste treatment process.

2.1.2.4 Process Control i The system is control 3ed by the UPMP process control system. (The UPMP process control system is described in Section 2.6.)

The determination of uranium t will be made from grab t samples which will be analyzed in the UPMP process control laboratory. The control room operator will be i notified of the results and, when t 1

will direct the control system t t

t i .

Periodic sampling and analysis of the product will be" '. '

made to insure that the process is operating within the limits. The limit of the effluent is t ppm uranium. If the concentration is above ppm t uranium, the UPMP process control system will j automatically switch flow to the surge tank V-106 until

the < ppm uranium is achieved. t i

t t

t

, t t

t

[ A proportional sampling system in the t

, effluent line will measure the total volume transferred i to waste treatment tank V-108 and will provide a j representative sample of the liquid transferred.

2.1.2.5 GECO Rundown System

' The source of fluoride waste for the existing GECO rundown tanks will be relocated from the existing surge tank discharge pump to a location downstream of the i

l l Page 2.1-9 References SNM-1097

+

June 1, 1984 1

process system. A controlled side stream of fluoride waste will be diverted from the pipe line to tank V-108, i then to the GECO rundown tanks A and B. The return from .

4 this system goes back into the transfer line to V-108. I This provides the rundown tanks with fluoride waste that has less than ppm uranium. t 2.1.2.6 Storage Tank - Existing Tank V-108 The treatment of the fluoride waste at the Waste Treatment Facility will not change significantly as a

result of the process system, except that uranium

i. bearing sludge will no longer be settled and 1

centrifuged. The reason for this is that the uranium

} concentration in the product is decreased by a factor of j or more from that in V-108 prior to the t implementation of the UPMP process.

The lime treatment of this fluoride waste to precipitate i fluoride and recover ammonia will not change, however, '

I the uranium concentration in the calcium fluoride sludge that settles in the fluoride lagoons will change.

Before implementation of fluoride waste treatment, the j uranium concentration in the dry calcium fluoride sludge i has.been about ppm. With the +

implementation of the treatment process, the !

concentration in the dry calcium fluoride sludge will be

about ppm or less and will make disposal of this t

material under SECY 81-576 an attractive option.

f 2.1.3 Criticality safety considerations

! i 1

As noted in Section 2.1.1, the fluoride waste system

will be designed to process an ammonium fluoride liquid

waste stream containing a very low concentration of uranium. Prior to the implementation of UPMP, this low

concentration has been in the range of ppm U in t

! the liquid stream and ppm U in the calcium t fluoride sludge in the final process lagoons. These values are factors of or more below the t minimum critical concentration for a homogeneous mixture of UO 2 and water. As indicated by the results in Section 4.1.9.1, the Kinf = 1.0 value is approximately 200,000 ppm U or, equivalently, about 250 grams U/ liter. <

i l

1 I

j Page 2.1-10 !

References SNM-1097 I L.~.-_ _ _ , - _ . , _ _ . _ _ _ _ . _ _ _ . _ . _ . . _ _ . _ _ _ _ . _ . . _ _ . _ _ _

2 June 1, 1984 i

The implementation of the UPMP system t

, between the existing fluoride waste system process tanks l i

V-106 and V-108 will favorably impact the fluoride waste l system concentration control as shown in Table 2.1-1.

i As indicated, no concentrations in the UPMP fluoride waste system will approach within less than a factor of i the 250 gm U/ liter minimum critical 002+HO 2 t concentration and only the concentrations in the product and in the '

t will be higher than those in the fluoride t waste system operation prior to UPMP.

i The following sections describe the specific controls implemented in the UPMP fluoride waste system to assure criticality safety. These controls include not only those existing prior to UPMP which assured concentration

control in V-106 and V-108 operations, but also the

process and chemistry considerations which limit j

achievable uranium concentrations and the extensive use;

! of geometry control in new UPMP fluoride waste '

equipment. A final section addresses the criticality safety of fluoride waste system under normal and accident conditions.

2.1.3.1 Density Control in UPMP Surge Tank - Existing V-106 i

Surge tank V-106 is a geometrically unsafe 65,000 gallon 4

tank located immediately outside of the Fuel

. _ . _ _ . _ Manufacturing Operations building. Tank V-106 is used as an accumulation tank for liquids from the fluoride

) -

quarantine tanks prior to their treatment by the UPMP i system. Criticality safety of the tank is based upon the uranium dump limit control of the quarantine tanks (in FMO) and adequate mixing of the contents of V-106 by a high volume liquid recirculation line. The fluoride j quarantine tank dump limits and V-106 mixing will not be

changed from those used by operations prior to UPMP. In j addition, it will be assured that not only are the j

specified concentrations not exceeded, but also that tank V-106 does not contain at any one time more than

, 75% of a minimum critical mass. (For 4.0% enriched UO

and considering the volume of tank V-106, this latter 2

! guideline is equivalent to a dump limit to V-106 of no

{ more than ppm U in each gallons released t l

i i

Page 2.1-11 Reference SNM-1097 l

k

- t i

f June 1, 1984 e i

f i

TABLE 2.1-1 !

FLUORIDE WASTE SYSTEM i URANIUM CONCENTRATION CONTROL i Uranium Concentration Range !

Prior to UPMP After UPMP Overall i System (ppm U) (ppm U) Effect !

i-Tank V-106 Unchanged t f

GECO Rundown Improved t [

4 Tanks i.

i p

t Recovery

t f t

Recovery

t l

)

Tank V-108 Improved .' l I

t Liquid Improved t ;

Discharge ;

f Calcium Improved t j Fluoride Sludge !

t I

f t !

t l i

i i i i

?

I i

f

\ r i

Page 2.1-12 [

Reference:

SNM-1097 t

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

June 1, 1984 from the quarantine tanks.) In addition, dissolution and the density monitoring feature will be added specifically for the UPMP operation. The dissolution of uranium and the density monitoring in tank V-106 in the UPMP operation will be predominately for process control, but will also serve to enhance the safety in V-106. The former (dissolution) will be an enhancement because it improves the mixing of the contents of V-106 in conjunction with the high volume recircu2ation line. 'The latter (density monitoring) will provide additional safety because a maximum density limit equivalent to grams U/ liter will be interlocked via t the UPMP process control system with the inlet valves to V-106. If this value (which is well below the minimum critical value for homogeneous UO, + H2 O) is reached, the process control system will close the V-106 inlet valves and not permit additional dumps from the fluoride quarantine tanks.

~ ~

2.1.3.2 Chemistry and Procers Control Chemistry t The basic chemistry of the process is t illustrated in Figure 2.1-4. The chemistry is divided into two cycles, t 2.1.3.2.1 t Fluoride liquid waste contains about pm t fluoride, ppm nitrate and ppm uranium. t t

t t

Page 2.1-13 Reference SNM-1097

June 1, 1984 FIGURE 2.1-4 BASIC CHEMISTRY t t

t t

t I

t t

i t

t t

't t

t t

t

.t t

t t

l l

Page 2,7_94

Reference:

SNM-1097

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4

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, . a u -

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s . n,e t e N- ' -

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, 't 2.1.3.2.3 Process Control .

,~~

The process is monitored and controlled by the UPMP j process control system in conjunction with the UPMP process' laboratory. This laboratory is located iu j adjacent to the UPMP control . room and contains spectrographic, titration and specific ion electrode instruments. JReference Figure 1-3.) iA, grab sa.$ple from the process will be analyzed for uranium and the data logged in the contspl room. The periodic!qrpb sample of the: process feed solution will be andlyzed in '

j the laboratory ~ for both soluble and insoluble uranium.

s v

i' l

$ t r

.. a- ,

s N ',

s - Page 2.1-15 \

Reference:

~

SNM-1097 _x t

e

4 June 1, 1984 FIGURE 2.1-5 t

t t

I t

t t

t

t t

t t

t f

Page 2.1-16

Reference:

SNM-1097

June 1, 1984 FIGURE 2.1-6 .,

t t

t

t t

I t

t t

t Page 2.1-17

Reference:

SNM-1097

June 1, 1984 FIGURE 2.1-7 t

t t

1*

t t

t t

t t

1 t

t t

1 t

t t

Page 2,j_gg i

Reference:

SNM-1097

~

June 1, 1984 This will determine if process adjustment are necessary for the dissolving of the uranium in surge tank V-106.

The UPMP process control system will modify process flows and chemistry when the process control samples indicate the predefined control limits have been reached. Control room operators will be aware of the system status.

The lab system will analyze the effluent from the process to insure that the uranium concentration is less than ppm. This system will provide automatic t >

sampling and analysis as well as manual capabilities and :

will be coupled with the process control system.

For the treatment system, a measurement greater than ppm uranium will result in the process control t

' system sounding an alarm and automatically diverting the effluent stream back into surge tank. The entire system will remain in this recirculating mode until the '.

discharge is determined to be below the ppm uranium t level, at which time the process control system will re-establish the flow to V-108.

A proportional sampling system will be located in the effluent discharge line from the treatment system to

tank V-108. This system will measure the flow rate and total volume discharged to tank V-108 and obtain a

- representative sample of the liquid flowing through the 1 pipe line. The sample will be analyzed daily by the !

Chemet laboratory for the uranium concentration and the results will be used for uranium accountaoility of the .

operation.

2.1.3.3 Geometry Control in the Fluoride Waste System In addition to the uranium density control described in the preceding sections, the UPMP fluoride waste system also makes significant use of geometry control for equipment in the treatment system. In summary, the 4

fluoride waste system consists of the following:

i L Page 2.1-19 L

Reference:

SNM-1097 l

June 1, 1984 (1) geometrically safe cylindrical i tanks.

(2) geometrically safe vessels. t (3) geometrically unsafe tanks. t (4) geometrically safe pumps, t (5) geometrically safe diaphragm pump. t (6) critically safe floor basin. t (7) Associated geometrically safe process piping '

Table 2.1-2 is a listing of the process equipment in the fluoride waste system and briefly describes the geometry and corresponding criticality 6afety designation. (HVAC equipment is not included in this list but is i, generically discussed in Section 5).

2.1.3.4 Normal and Accident Conditions in the Fluoride Waste system As noted previously, the maximum uranium concentration i present in the fluoride waste system is grams per i l

liter. From Table 4-4 in Chapter 4, the Kinf for a full density homogeneous mixture of UO and H O with a 0

-density no greater than this is in 2the range 2 of 0.30-0.50. It can therefore be concluded that under normal conditions, the Keff for any piece of equipment in the fluoride waste system is well below a value of 0.90.

In the evaluation of fluoride waste system accident conditions, four significant conditions have been identified. These are:

2.1.3.4.1 Failure of Concentration Control in Existing Tank V-106 In order for a significant failure of concentration control in existing surge tank V-106 to occur, at least l

Page 2.1-20

Reference:

SNM-1097

1

\

r June 1, 1984 i

three independent system failures must occur upsteam prior to dumping from the fluoride quarantine tanks.

l (1) Failure of manufacturing process equipment with a i

subsequent high carry over of uranium into the fluoride liquid waste system.

(2) Failure of fluoride quarantine tank process controls (most notably the filter and centrifuge systems) to remove the high concentration of uranium from the ;

stream prior to transfer to the quarantine tanks. '

(3) Failure of the U monitor or SPEC 20 measurement systems to detect the high level of uranium prior to dumping of the quarantine tank to the surge tank.

Given these failures and the following assumptions about credible conditions, an assessment of the criticality safety of tank V-106 has been made. -

o Assume a single 1000 gallon quarantine transfer containing as much as 100 kg of uranium is dumped to V-106.

o Assume that prior to this dump, tank V-106 contained 20,000 gallons of liquid (approximately 1/3 full) with approximately ppm U. t o Assume the mixing, dissolving and density controls on tank V-106 are fully operable and monitored by the UPMP process control system.

i !

Based upon these conditions, it can be seen that after the kg U per gallon quarantine transfer is t dumped to tank V-106, the surge tank will contain l kg U and gallons of liquid / mixture. t This mixture will be adequately dissolved and mixed by process controls, and the resulting uranium

{ concentration will be no more than grams of t

. uranium.per liter. Since quarantine tank dumps are only made_approximately every 30 minutes, continued releases to V-106 at this level for more than a few times are not credible. Nonetheless, if such releases were to occur over_a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the resulting slow increase,

, Page 2.1-21

Reference:

SNM-1097

, . - _ . _ . _ _ _ . _ _ _ , _ . . - __ _ ,__._ _ _._.___._..____.._.__...__i

b June 1, 1984 l

in uranium content in V-106 would eventually be detected !

by the surge tank density controls and further input i from the quarantine tanks halted by the process control i system. (At least such dumps would be required to t !

reach this point since the V-106 density limit is set at !

grams U/ liter). t Because at least four independent and multiple failures 4 must occur to exceed this critically safe condition, this system complies with the double contingency policy defined in Chapter 4. ;

I 2.1.3.4.2 Failure of Concentration Control in the System !

i Except as noted in Table 2.1-2, all process equipment in {

the treatment system is geometrically safe. Failure of j concentration control in this area will therefore not t result in a critically unsafe condition without the i i

coincident and unrelated failure of such a geometry - ;

control.

Conversely, failure of geometry control in the treatment f system will not result in a critically unsate condition ;

without the coincident and unrelated failure of the l fluoride waste system concentration controls. ; '

s 2.1.3.4.3 Back Flow of Uranium Into the DI Water Supply Tank or ,

Make-up Tank i i

Since the treatment system is operated under 1 i

lbs positive pressure, prevention of backflow of uranium into the unsafe geometry DI H 2 O supply tank and i

make-up-tank is achieved by process controls (monitoring !

l devices) and automatic double block and bleed valves. i l The double block and bleed valve operation is shown in i Figure 2.1-8. !

! i l If the process control and double block and bleed valve :

systems do fail, however, criticality safety will still l l be assured by the fluoride waste system concentration i controls. Not only will the uranium concentration i backflowing into the DI water supply tank and the !

make-up tank be further diluted from the value in the [

fluoride waste system, but also-the maximum volumes of j i

t I

y i

, Page 2.1-22 i

Reference:

SNM-1097

June 1, 1984 TABLE 2.1-2 UPMP FLUORIDE WASTE SYSTEM EQUIPMENT LIST Normally Equipment Contains Criticality Safety Uranium Geometry Description Description Label Equipment Designation Treatment Equipment Yes Geometrically Safe with t

Neutron Absorber Panels t t

Pilters Yes Geometrically safe Cylinders t t

t Filters No, Geometrically safe Cylinders t ppa U t t

Filters Yes Geometrically Safe Cylinders t t

t venturi Gas Scrubber Yes Geometrically Unsafe Tank but t part of T-500 Operation t t

Pump Yes Geometrically Safe Slab s t volume ,

, t Pumps Yes Geometrically Safe slab 6' t Volume t t

Pump Yes Geometrically safe Slab & t volume t Pumps Yes Geometrica1' Safe Slab & t Volume t t

Pumps Yes Geometricahiy Safe Slab &

Volume t t

Pump No Geometrically Safe Slab and ?

Volume t t

Seal Pot Yes Geometrically Safe cylinder t!

ti tl Seal Pot Yes Geometrically Safe Cylinder t; t;

rw Sump No Geometrically Safe Cylinder t (with Concrete Reflection) t t

Break Tank No Geometrically Safe Cylinder t t

t Break Tank No Geometrically Safe Cylinder t t

t Tank Yes Geometrically Unsafe t

' t t

! Tank Yes Geometrically Safe Cylinders t t

l t j D1 Water Supply Tank No Geometrically Unsafe Back-flow t I

prevented by Double Block & t Bleed Valves t t

Page 2.1-23

Reference:

SNM-1097

June 1, 1984 TABLE 2.1-2

/ UPMP FLUORIDE WASTE SYSTEM EQUIPMENT LIST (CONTINUED)

Normally Equipment - Contains Criticality Safety Label Equipment Designation Uranium Geometry Description Description t

Tank No Geometrically Unsafe Back-flow t prevented by Double Block 6 Bleed Valve t

Tanks No, Geometrically Unsafe i ppa U t t

Tank No, Unsafe Geometry t ppm U t t

vessels Yes Geometrically Safe with . t Neutron Absorber Panels t t

t vessels Yes Geometrically Safe with t Neutron Absorber Panels Process Piping Yes 1*,2*, 3* and up to 6* Geometrically Safe Cylindrical diameter piping and Slab Geometries PW Floor Basin No 3.5* Thick Slab Critically safe Slab with 'g Concrete Reflection Page 2.1-24

Reference:

SNM-1097

June 1, 1984 FIGURE 2.1-8 BLOCK & BLEED VALVE SYSTEM F

E -

N /' N /' m p tow A A '

1 2 The system is controlled by a sequence controller that ;

is activated by an OPEN or CLOSE signal.

NORMAL FLOW POSITIONS Block Valve #1 is OPEN Block Valve #2 is OPEN Bleed Valve #3 is CLOSED VALVE OPERATION CLOSE:

To stop the flow, the CLOSE signal is activated. The valves operate in sequence. That_is, when block valve

1 closes, then block valve #2 closes, then bleed valve 43 opens. If either block valve #1 or 2 should leak, the liquid will flow out through the bleed valve #3 onto the floor.

OPEN:

To open the valves,'to start the flow, the OPEN signal is activated. The valves operate in sequence. That is, when bleed valve #3 closes, then block valve #2 opens, then block valve #1 opens.

Page 2.1-25

Reference:

SNM-1097

June 1, 1984 t

these tanks assures that the amount of uranium contained !

in either of the tanks will be less than a safe batch (as tabulated in Table 4-7).

2.1.3.4.4 Failure of Concentration Control in V-108 [

Failure of concentration control in tank V-108 will !

require as a minimum the following independent, '

unrelated conditions:

c (1) Failure of fluoride quarantine tank dump limit l controls. i,

~

i (2) Failure of treatment process controls to remove

uranium. '

(3) Failure of the UPMP measurement system to detect l high ( ppm U) uranium concentrations. t i Since tank V-108 is itself equipped with concentration I controls and a solids removal system, failure of ,

concentration control will not violate the requirements i of the double contingency policy. !

8 i

- a t

i l

I I

Page 2.1-26 l

Reference:

SNM-1097 i

June 1, 1984 2.2 RAD WASTE TREATMENT 2.2.1 Current Operation The rad waste system is an accumulation of various low level radioactive waste streams that are generated in the fuel manufacturing operation. These streams, as shown in Figure 2.2-1, are collected from the Chemet laboratory, the decon facility, the incinerator, equipment cleaning for maintenance and general housecleaning liquids. These liquids are collected in the existing FMO rad waste accumulation tanks II and #2 which are located in the FMO rad waste complex. The pH is frequently adjusted by the operator to reduce the soluble uranium content of the rad waste. The liquids are circulated through an existing clarifier to continuously remove suspended uranium and other particulate.

The solids removed by the clarifier are backflushed to !

an accumulator tank. The slurry is desludged using a dewatering, bowl-type centrifuge. The solids are loaded into containers, transported to the REDCAP area for drying and subsequently shipped off-site for processing.

The reprocessed material is returned to WMD as UNH solution.

I The clarified rad waste is transferred to the existing FMO rad waste slab tanks 4 and 5. Laundry water is also transferred from its collection point in the existing FMO-X rad waste slab tanks to FMO rad waste slab tanks.

The composite liquids in the tanks are circulated, sampled, and analyzed for uranium. If the uranium concentration is less than the release limits ,

of ppm U, the liquid is discharged to the existing t

. industrial sewer. The discharge line is equipped with a continuous proportional sampler for uranium accountability. The uranium in the rad waste is precipitated and is settled in the existing final process lagoons after it mixes with the other waste i- streams in the industrial sewer. The uranium concentration of this dried sludge ranges from ' t ppm. t l

I l

Page 2.2-1

Reference:

SNM-1097

June 1, 1984 4 FIGURE 2.2-1 RAD WASTE - PRESENT FLOOR LhB DECON CLEANING if if ir INCINERATOR > SUMP sr RAD WASTE Slab Tanks if ;

i CENTRIFUGE > SLUDGE t'

1/

QUARANTINE 4-- LAUNDRY WATER 1 Slab Tanks 1 y

ppm U t PROCESS DRAIN 3/~

FINAL PROCESS LAGOONS l 3r LIQUID EFFLUENT < 1 ppm U i SLUDGE ppm U- RIVER t d

I i<

i Page 2.2-2

Reference:

- SNM-1097

.- = . - . .-. _ _

June 1, 1984

, 2.2.2 UPMP Radwaste Treatment System The UPMP rad waste treatment system will be implemented ;

to improve the uranium recovery by:

o Providing a controlled lime treatment process.

e Providing a filtration system for liquid an5 solids separation.

The results will be a substantial reduction of uranium concentration in the sludge settled in the final process lagoons.

2.2.2.1 Rad Waste Accumulation In a concept similar to the UPMP fluoride waste system, the UPMP rad waste system will be situated between the

~

existing rad waste quarancine system and the current ;

discharge point to the final process lagoons. (See Figure 2.2-2) Several of the waste streams presently being collected in the existing FMO rad waste accumulation tank will be diverted to the UPMP rad waste accumulation vessels. The only waste streams that will be collected in the present FMO rad waste accumulation

slab tanks will be the laboratory waste water, equipment cleaning water and housecleaning liquids. These waste liquids will continue to be treated by the present process, accumulated in the FMO rad waste slab tanks, analyzed, released, and pumped to the UPMP accumulation vessels. The existing releane limits of ppm t uranium will still be in ef fect.

2.2.2.2 Laundry Water The laundry water will continue to be collected in the existing FMO-X slab tanks. When one tank is full, the input water will be diverted to the second tank. The water in the full first tank will .be circulated for mixing, pumped through the sample loop to UPMP laboratory, and sampled and analyzed for uranium by the UPMP lab measurement system. The results will be transmitted to the UPMP process control system which will automatically control the dispositon of the laundry Page 2.2-3 l

Reference:

SNM-1097

=-

June 1, 1984 FIGURE 2.2-2 RAD WASTE - UPMP LAB v

MISCELLANEOUS --4> SUMP t--- FLOOR CLE ANING h

RAD WASTE Slab Tanks 5

CENTRIFUGE --4> SLUDGE h

RAD WASTE Slab Tanks 5

. DECON --4m ACCUMULATION 4--- INCINERATOR i

> VESSELS 4--- MISCELLANEOUS '

4 R-605 4--- LIME Pipe Tank REACTOR <

4

{

TANKS 4r FILTERS ; CENTRIFUGE 4

V RECEIVER VESSEL SLUDGE gg Key:

LAUNDRY -4> TANK UPMP WATER o

PROCESS ---) -

FINAL PROCESS DRAIN LAGOON ppm U i t

RIVER Effluent < 1 ppm U ppm U t l

Page 2.2-4

Reference:

SNM-1097 l

June 1, 1984 i

i FIGURE 2.2-2 RAD WASTE - UPMP (CONTINUED) t t

t

. t ,

t ,

t t '

t t

t t

t t

t i t 1

t t

t ll t t

t t

. t t

t i

I l

1 l

Page 2.2-5 l l

l

Reference:

SNM-1097 i l

l

) ,

l s

June 1, 1984 l

water. If the uranium concentration is less than a preset value ( ppm), the water will by-pass the t treatment facility and be pumped to the rad waste discharge point. If the uranium concentration is above the preset value it will be pumped to the accumulation vessels for normal treatment.

2.2.2.3 Rad Waste Accumulation Vessels The waste' water in the accumulation vessels will be kept in constant circulation. A controlled side stream will be diverted to the tank where it will be continuously mixed with a % lime. slurry to a pH of and then overflow into the holding tanks. .-

The lime addition will precipitate fluorides, uranium and other ions as calcium compounds.

The % lime slurry will be supplied to the UPMP ; t facility from a gallon storage tank which is '

t located outside the west end of the building in the aqueous and solvent make up area. The lime slurry will be made up on a batch basis, by diluting % lime t slurry from a gallon storage tank with rad waste t from the the rad waste holding tank. +=

l l 2.2.2.4 Rad Waste Tanks t

-The lime treated water solution will be a dilute slurry as it enters the tanks. Thorough mixing by circulation will prevent settling of any sludge in the t vessels. The dilute slurry will be pumped from the rad waste vessels into the filter system.

2.2.2.5 t t

i t t

t t

Page 2.2-6

Reference:

SNM-1097

l June 1, 1984 i FIGURE 2.2-3 t

t t

t

i t

t t

i t

i l

Page 2.2-7

Reference:

SNM-1097

June 1, 1984 t

t t

t t

t t

t, 1

t 2.2.2.6 t t

1 t

t t

t 2.2.2.7 t '

t t

t l Page 2.2-8

Reference:

SNM-1097

June 1, 1984 t

t t

t 2.2.2.8 t t

t t

. ; t t

t t

t 2.2.2.9 Rad Waste t The product, flowing by gravity into the receiver vessel, will be clear and contain less than t ppm uranium. The receiver vessel will be kept in constant circulation and a small side stream will be diverted through a constant flowing sample loop to the process control laboratory. There a sample from the loop will be withdrawn and analyzed for uranium by the UPMP lab system. The results will be transmitted to the UPMP process control system.

If the uranium is less than ppm, the product will be t continuously discharged to a gallon holding tank t located outside the west end of the building. If the uranium is grecter than ppm, an alarm will be t sounded in the control room and the product will be automatically diverted by the UPMP process control Page 2.2-9

Reference:

SNM-1097

June 1, 1984 system, back to rad waste collecting vessels for reprocessing.

The product collected in the holding tank will be discharged by gravity through a proportional sampler to the process drain. The level in the holding tank will be maintained at about 50% to provide dilution water to make up % lime slurry from % lime t slurry. ,

2.2.3 Criticality Safety Considerations The rad waste system is designed to process rad waste liquids containing low concentration of uranium, as '

noted in Section 2.2.2. Prior to the implementation of UPMP, the uranium concentration in this rad waste stream has been low, ( ppm or less). The sludge that is t removed from the rad waste stream prior to discharge has a uranium concentration of ppm. ; t These sludges are collected in five gallon pails, dried

and stored for off-site processing. The sludge that is precipitated in the final process lagoons has a low uranium concentration of ppm. These t values are a factor of or more below the t minimum critical concentration for a homogeneous mixture ,

of pure UO2 and water. As indicated by the results in l Section 4.1.9.1, the Kinf = 1.0 value is approximately 200,000 ppm U or equivalent to 250 grams uranium / liter.

The implementation of the UPMP rad waste system between the rad waste quarantine tanks and the discharge to the final process lagoons will favorably impact the rad waste system concentration control as shown in Table 2.2-1. The major impact-will be reduction of uranium concentration in the liquid discharge by a factor of t or more. t The following sections describe the specific controls implemented in the UPMP rad waste treatment system to assure criticality safety. These controls include not only those existing prior to UPMP, but also the process considerations which limit achievable uranium concentrations. Also, extensive ut - of geometry in the l

Page 2.2-10

Reference:

SNM-1097 l _. _ -_ __ __ - -

June 1, 1984 TABLE 2.2-1 RAD WASTE SYSTEM URANIUM CONCENTRATION CONTROL Uranium Concentration Range, ppm U Prior After Overall System to UPMP UPMP Effect Discharge from Unchanged t Rad Waste Quarantine Tank

Vortex Clarifier Unchanged t Sludge t Rad Waste Surge t Accumulation Capacity *.

Tank

Rad Waste Pre- t Tanks Treatment

t Filters Recovery

t

?

Centrifugal Sludge Recovery

t

_ . _ . . t Rad Waste Product Improved t Discharge to Improved t Final Process Lagoons Sludge in Improved t Final Process t Lagoons

Provides for direct recovery and internal LLcycle of the ;

uranium.

t i

a Page 2.2-11

Reference:

SNM-1097

. June 1, 1984 UPMP rad waste treatment equipment assures criticality safety. A final Section 2.2.3.4 is included which discusses the criticality safety of the rad waste system under normal and accident conditions.

2.2.3.1 Concentration Control in UPMP Rad Waste Accumulation Vessels The operation of the rad waste accumulation vessels is detailed in Section 2.2.2. These vessels are in design with a and are t geometrically safe by virtue of t neutron absorber panels.

l After the implementation of the UPMP rad waste treatment system, several of the lowest uranium concentration waste streams will be diverted from the FMO rad waste quarantine tank directly to the UPMP rad waste accumulation tanks. The waste stream that is the major; contributor of uranium (equipment cleaning for maintenance), will remain in the FMO rad waste !

quarantine system. This waste stream will continue to be processed through a clarifier desludging operation prior to accumulation in the existing rad waste quarantine slab tanks. The existing rad waste ,

quarantine tank dump limits are unchanged and the 1 <

uranium concentration will be verified by a SPEC 20 measurement to be below ppm. This measurement t will assure that the uranium sludge is removed before the rad waste is transferred to UPMP accumulation!

vessels.

Laundry water will not normally be routed to the rad waste accumulation vessels. It will normally by-pass' '

' the treatment system and be discharged through the proportional sampler to the industrial sewer. The laundry water will be accumulated in the existing slab tanks, circulated and analyzed by the UPMP laboratory systems. The uranium content in the laundry water very seldom reaches as high as ppm. If the uranium t

, content is below .

ppm, the water will be by- t passed around the treatment system. Consequently, when laundry water is routed through accumulation vessels, it will actually cause additional dilution of the uranium, i

! I l

l Page 2.2-12 j

Reference:

SNM-1097 j l

June 1, 1984 In addition to these concentration controls, the rad waste accumulation vessels have also been instrumented with density monitors interfaced to the UPMP process control system. If these density monitors detect values higher than the equivalent of grams uranium / liter, t the process control system will automatically close all process inlet valves end sound an alarm to notify the control room operator of the condition.

2.2.3.2 Rad Waste Treatment Chemistry and Process Control 2.2.3.2.1 Process Chemistry The rad waste accumulated in vessels V-600 and V-601 will contain soluble uranium, carbonates, fluorides and numerous metal ions. A controlled stream of rad waste flows from the accumulation vessels to the tank. The liquid in the tank will be thoroughly mixed with an agitator. A % lime slurry will be added to . t the tank to adjust the pH to about This lime

t will precipitate the uranium, carbonates and fluorides as calcium compounds and the metal ions as hydroxides.

The pH of the rad waste stream is measured before and after the tank, for control of the addition of % t lime slurry to attain the proper pH. The treated rad waste from the tank will flow into the rad waste vesuels to stabilize the reaction. t 2.2.3.2.2 Process Control P

5 The UPMP rad waste treatment system will be controlled

by the UPMP process control system. The flow rate ,

through the process will be controlled by the liquid 4

level in the radwaste accumulation vessels. This level is dependent on the liquid input from the various j sources of rad waste. t t

t

t t

t t

t I

l Page 2.2-13

Reference:

SNM-1097 l

4 June 1, 1984 t

t t

t i t

t t

i The pH of the product in the pipe line header will be monitored by the UPMP process controller. Any change in '

product pH will be noted by the controller. A pH condition outside the operating parameter range will sound an alarm.

The product in the receiver vessel will be circulated- ;

, and the discharge rate from the system controlled by the' liquid level in the vessel. A small side stream will be diverted through a constant flowing sample loop to the UPMP control laboratory. There a sample will be withdrawn from the sample loop and analyzed for uranium by the UPMP analytical system.- The results will be ;

transmitted to the UPMP process controller. If the I uranium is less than ppm, the product will be t-

, continuously discharged to the gallon holding t i tank. If'the uranium is above ppm, an alarm will be t sounded and the product automatically diverted back to the radwaste accumulation vessels for reprocessing.

The density of the product will be monitored by a density meter in the receiver recirculation system.

~ If

this density should exceed the equivalent of grams t uranium per liter, an alarm will be sounded and the product stream automatically diverted back to receiver vessels for reprocessing.

Water that-is discharged from the laundry facility will be collected -in existing geometrically safe slab tanks, circulated, sampled and analyzed for uranium by the UPMP laboratory system and the results transmitted to the UPMP process controller. If the-uranium concentration i

Page 2.2-14

Reference:

SNM-1097

June 1, 1984 is ppm or less, the laundry water will by-pass t the rad waste treatment and be discharged through the proportional sampler, to the process drain.

If the uranium concentration is above the preset value, the laundry water will be discharged to the rad waste accumulation vessels to be treated.

The product from the rad waste product receiver vessel will b discharged into the product holding tank located i outside the west end of the building. The liquid level l l

in this tank is maintained at about 50% to provide dilution water for the % lime slurry make-up. t Any liquid above the 50% level will be discharged through a proportional sampler to the process drain.

The proportional sampler will measure the volume discharged and obtain a representative sample of the liquid for uranium analysis by the WMD Chemet ;

laboratory. '

2.2.3.3 Geometry Control of the Rad Waste Treatment System The rad waste treatment system makes significant use of geometry control as well as the use of uranium concentration control described in the preceding sections. In summary, the rad waste treatment system consists of the following:

(1) geometrically safe cylindrical tanks. t (2) geometrically safe vessels. t (3) geometrically unsafe tanks. t (4) existing geometrically safe slab tanks. t (5) geometrically safe filters. t (6) geometrically safe cylindrical t centrifuges.

Page 2.2-15

Reference:

SNM-1097

June 1, 1984 i

\,

i (7) geometrically safe pumps. ,

I (8) geometrically safe diaphragm pumps. t

~

(9) One (1) critically safe floor basin for liquid i spills.

l l

(10) Associated geometrically safe process piping. !

(11) Material handling equipment in compliance with geometrically safe designs: five (5) gallon pails.

Table 2.2-2 summarizes the process equipment along with '

) a brief description of the geometry and the -

corresponding criticality safety designation. (HVAC is i not included in this listing but is generically i discussed in Chapter 5). !

. 2.2.3.4 Normal and Accident Conditions in the Rad Waste System ;

i As noted previously, the maximum uranium concentration I normally present in the UPMP rad waste system is less ,

j than grams / liter. From Table 4-4 (in Chapter 4), t '

the Kinf.for a full ~ density homogeneous mixture of UO and H 2 O with a U density no greater than this is in t$e f range of 0.30-0.50. It can, therefore, be inferred that } i under normal conditions, the Keff for any piece of [

equipment in the rad waste system will be well below a i value of 0.90. !

In the evaluation of the rad waste system accident conditions, the following three significant conditions ,

j have been identified: !

i 2.2.3.4.1 Failure of Concentration Control in the Rad Waste

System i

l i

i As noted in Table 2.2-2, almost all process equipment in the UPMP rad waste system is geometrically safe and i failure of rad waste system concentration control will l

not result in'a critically unsafe system. The only ;

exceptions to this are product holdup tanks and the !

lime storage tanks. The latter two of these are large

geometric process support vessels not designed to ,

t f

f f

l l i

I Page 2.2-16 i

Reference:

SNM-1097 .

L f n :

June 1, 1984 s TABLE 2.2-2 s.

UPMP RAD WASTE SYSTEM EQUIPMENT LIST Normally =

Equipment Contains Criticality Safety Label Equipment Designation Uranium Geometry Description Description

' Centrifuges Yes Geometrically Safe Cylinder t t

t P11ters Yes Geometrically Safe t t

t Pumps Yes Geometrically Safe Slab & t volume t t

Pumps Yes Geometrically Safe slab & t volume t t

Pumps Yes Geometrically Safe volume t

t Pumps No, Geometrically Safe Slab and t ppm U Volume * , '

t

~ - t

~

Pump No, Geometrically Safe Slab 6' t ppm U Volume t t

Pump No, Geometrically Safe Slab & t ppm U Volume t t

Pumps Yes Geometrically Safe Slab and i Volume t t

Pump No Geometrically Safe Slab and i Volume t t

Tank Yes Geometrically Safe Cylinder t t

t Seal Pots No Geometrically Safe Cylinder t t

t Sump No Geometrically Safe cylinder t t

Tanks Yes Geometrically Safe Slab t t

t Tank No, Geometrically Unsafe t Ppm U t t

Tanns Yes Geometrically Safe cylinder t t

Tank go Geometrically Unsafe t t

t No Geometrically Unsafe Tank t

I l

l l

l Page 2.2-17

Reference:

SNM-1097

June 1, 1984 l TABLE 2.2-2 UPMP RAD WASTE SYSTEM EQUIPMENT LIST (CONTINUED)

Normally Equipment Contains Criticality Safety Label Equipment Designation Uranium Geometry Description Description vessels Yes Georetrically Safe with t Neutron Absorber Panels t t

vessels Yes Geometrically Safe with t Neutron ADsorber Panels t'

Vessel No, Geometrically Safe with t ppm 0 Neutron Absorber Panels t t

Process Piping Yes 1*,2*,3*,4*,5*,8* Pipe Geometrically Safe Cylinders &

Slabs Rad Waste Floor Basin No 3.5" Thick Slab CriticallySafeSlabwith[

Concrete and 1* Water Reflection rive Gallon Pails at Yes 11-1/4* If, 13-1/2* High, 22 Geometrically Safe Volume CEM-644 6 646 Discharge Liters 1

Page 2.2-18

Reference:

SNM-1097 i

June 1, 1984 contain uranium and are discussed further in Section i 2.2.3.4.3. 1 l

Failure of concentration control in hold-up tank !

requires conditions:

the following independent arig coincident (1) Failure of concentration control in the tad waste quarantine tank system including mot only failure to restrict high level uranium straams from the system but also failure to detect the high U content via the SPEC 20 colorimeter.

(2) Failure of the UPMP (geometrically safe) rad waste treatment process especially the precipitation and filtration steps to remove the high level of uranium in the stream.

(3) Failure of the UPMP process control system to .

detect the increase of the turbidity in the .

product from the filters; failure of the density monitors on the product receiver vessel; and failure of the process laboratory system to detect the high uranium concentration, and the failure of the pH control system.

For an unsafe condition to arise, all three of the independent cases must occur at the same time. It is therefore concluded that the criticality safety of the product hold-up tank is fully in compliance with the double contingency policy.

2.2.3.4.2 Failure of Geometry Control in the Rad Waste System Failure of geometry control in the rad waste system will not result in a critically unsafe condition without the coincident and unrelated failure of the rad waste concentration controls.

2.2.3.4.3 Backflow of Uranium into the Lime Slurry Tanks Backflow of uranium into the lime slurry tanks from the rad waste system is prevented by the use of a positive Page 2.2-19

Reference:

SNM-1097 l

June 1, 1984 air break in the process piping from the lime slurry tanks to the rad waste system. This is achieved by piping the lime slurry solution into the top of the rad waste reactor tank which is vented to the atmosphere at a point above the solution level in the tank.

If this air break backflow prevention fails, however, the lime slurry tanks will still be critically safe because of the rad waste system concentration control.

e 1

~ i.

1 l

l l

1 Page 2.2-20

Reference:

SNM-1097

Juns 1, 1984

'2.3 NITRATE WASTE TREATMENT 2.3.1 Current Operation Nitrate waste is an accumulation of radioactive waste streams that are gen rated from the existing Uranium Purification Process (UPS), from acid flushing of the existing ADU conversion lines and from acid cleaning of equipment for maintenance. The UPS conversion waste stream'is collected in the FMO nitrate waste quarantine tanks, analyzed and transferred to accumulation tank V-103 if the uranium concentration is ppm or less, t If the uranium concentration is between t ppm, the quarantine tanks may be dumped to V-103 with authorization by the foreman. If the quarantine tank concentration is greater than ppm, the material t must be recirculated / reworked until it is within limits.

Figure 2.3-1 outlines the treatment process.

Nitrate Waste Quarantine

~ 3 2.3.1.1 The waste stream from the acid flushing and cleaning operation is collected in FMO nitrate waste C quarantine tank. When full, the liquid is circulated by a pump, ammonium hydroxide is added to precipitate any soluble uranium and the liquid is passed through a centrifuge to remove the solids. When the uranium concentration in the quarantine tank is ppm or less, the contents t

- -~

are transferred to accumulation tank V-103. The sludge removed by the centrifuge is placed in five gallon pails to be dried and shipped off-site for processing. The uranium concentration of the wet sludge is t ppm. t The liquid in tank V-103 is kept in constant circulation by a pump to prevent settling of uranium. A circulating stream is monitored for density and if the density exceeds a value equivalent to grams uranium liter t in the waste stream, an alarm is sounded in the control room and the nitrate waste inlet valves are closed.

Page 2.3-1

Reference:

SNM-1097.

A June 1, 1984 FIGURE 2.3-1 NITRATE WASTE TREATMENT PRESENT

' FLUSH '

UPS COtTVERSION AC::D NHnOH M::SC.

V it if V'

-QUARANTINE QUARANTINE

TANKS TANK & _

O CENTRIFUGE - i V

SLUDGE gg ppm U '

t TANK < l LIME if SLURRY --9 TANK > NITRATE

--# LAGOON CENTRIFUGE V TANKER-

- 3r TO OFFSITE USE 1 SLUDGE I

t ppm 0 -

L

=-

-~.

I

, p.

Page 2.3-2'

Reference:

-SNM-1097 1

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

, , - , . , , , , , . . , , - . . - , , -,,---.-.,,n,,-,,nw,,,,_. , , - - -

June 1, 1984 s

P'

~

'2.3.1.2 Nitrate Wast'e Treatment

. - The liquids that have been accumulated in tank V-103 are

"~ pumped, on demand, to treatment tank V-104 located at the Waste Treatment Facility. The liquid is batch

,< treated with lime to adjust the pH. Uranium and other i metal ions are precipitated. The batch is allowed to settle for hours, in order to decompose the t hydrogen peroxide used in the UPS conversion process.

^

After the solids are settled, the clear liquor is

~

decanted to the nitrate lagoon. The slu6ge is removed from the bottom of.the tank V-104 and centrifuged to remove the solids. The solids are placed in five gallon pails, dried and shipped off-site for processing. The uranium ~ concentration of this wet sludge is t

' ppm. The centrate is returned to tank t V-104. When the sludge removal is finished, the treatment' cycle is complete and ready to receive the .

next batch from tank V-103.

The' decant pipe line from tank V-104 to the nitrate lagoon is equipped with a proportional sampler for i

uranium accountability. The uranium concentration in the clear nitrate waste solution in the lagoons is about ppm and in the nitrate waste sludge is t ppm. t 1

, 2.3.2 UPMP Nitrate Waste Treatment Because of the installation of solvent extraction with UPMP and the modification of the UNH conversion process (old UPS conversion), the use of hydrogen peroxide will no longer be required. Thus, the nitrate waste treatment system can be changed. This waste treatment i . system will be installed between the existing quarantine tanks and the' existing Waste Treatment Facility. Figure 2.3-2 outlines the treatment process.

The treatment of the nitrate waste discharged from the various processes in UPMP and UNH- conversion will be segregated into two steps. The first step,-primary treatment, will treat all waste liquids except the aqueous waste (AW or Raffinate) from solvent extraction and will be very similiar in form to the rad waste l

Page 2.3-3

Reference:

SNM-1097

June 1, 1984 FIGURE 2.3-2 NITRATE WASTE TREATMENT UPMP 1

FLUSH UNH CONVERSION ACID NH nOH MISC.

if %/ i/ V QUARANTINE QUARANTINE TANKS- TANK d--

iir 4 ACCUMULATION < CENTRIFUGE TANK (--- MISCELLANEOUS 4 '(

---A REACTOR 4-- LIME SLURRY SLUDGE TANK

@ S/ '

CENTRIFUGE -

FILTER l

TANK SLUDGE y ppm 0 tl NITRATE WASTE SX RAFPINATE ---b STORAGE TANK

, REACTOR (--- LIME SLURRY

+

t- - b TANK --> NITRATE I V-104 LAGOON u 3r ppm U t F-440 TANKER FILTER TO OFFSITE USE Key: V

. SLUDGE UPMPl ppm U t l

l i

Fage 2.3-4

Reference:

SNM-1097 l

June 1, 1984 FIGURE 2.3-2 NITRATE WASTE TREATMENT (CONTINUED) t t

t t

t t

t t

I t

t t

l l

Page 2.3-5

Reference:

SNM-1097

June 1, 1984 operation. The second step, secondary treatment, will treat the aqueous waste (AW) from the solvent extraction process in addition to the primary nitrate waste product stream. The aqueous waste (AW) waste stream will contain all the metal impurities rejected by the solvent extraction process and will contain trace levels of uranium ( ppm). t The treated waste stream from the primary treatment system'and the aqueous waste (AW) will be blended in the existing nitrate waste storage tank V-103 and then treated in the secondary nitrate waste treatment located at the Waste Treatment Facility. The effluent frc..I this treatment will be discharged to the nitrate lagoon and -

the solids shipped to an approved off-site waste burial facility.

2.3.2.1 Primary Nitrate Waste Treatment 2.3.2.1.1 Nitrate Waste Collection Tank The nitrate waste from UNH conversion quarantine tanks, A and B, located in FMO, will.be pumped to the

! collection vessel. Other nitrate waste liquors generated in the UPMP facility, except the aqueous i waste, will also be collected in this vessel. The I liquid in this tank will be kept in constant circulation by the nitrate waste transfer pump to thoroughly mix and prevent settling of solids. This solution will have a density of gms/ml and a uranium concentration of t l about ppm. If the fluoride waste product is also t i being treated, the uranium concentration of the solution can be as high as ppm. t 2.3.2.1.2 Nitrate Waste Treatment A controlled stream of nitrate waste is diverted from the recirculation of the nitrate waste collecting vessel to nitrate waste reactor. In the reactor, a controlled

. volume of 4 lime. slurry is added and mixed with the t nitrate waste to maintain a pH of about The t treated solution, a dilute slurry, will flow into a vessel. This solution will be circulated by the filter pumps and thoroughly mixed to insere completion of the precipitation reaction.

, Page 2.3-G

Reference:

SNM-1097

June 1, 1984 2.3.2.1.3 t t

t t

t t

t t

t

[ t l t' t

2.3.2.1.4 t l t t

t l t i t

( t i t l

l l

l l Page 2.3-7

Reference:

SNM-1097

June 1, 1984 FIGURE 2.3-3 t

t t

t

t t'

t t

'Ti t

t t

t t <

t (Duplicate of Fig .c e 2,2-3) t t

Page 2.3-8

Reference:

SNM-1097

)

June 1, 1984 t

t t

2.3.2.1.5 t t

t t

t l t t

t t

t

, t t

t t

2.3.2.1.6 Nitrate Waste Product The product from the treatment system will flow by t

gravity into the product receiver tanks. These are diameter Schedule cylindrical tanks, t high. The liquid in these tanks will be kept in constant circulation by a product pump. A small side stream will be diverted through a constant flowing sample loop to the UPMP process control laboratory. l There, a sample from the loop will be automatically l withdrawn and analyzed for uranium. The results will be transmitted to the UPMP process control system. If the uranium is ppm, the product will be continuously t discharged to nitrate waste storage tank V-103 located outside of the building. If the uranium is greater than ppm, an alarm will be sounded and the product will t Page 2.3-9

Reference:

SNM-1097

l June 1, 1984 l

l be automatically diverted by the UPMP process control 4

system, back to nitrate waste collecting vessel for t reprocessing.

2.3.2.1.7 Nitrate Waste Storage Tank - Existing V-103 V-103 is an existing 20,000 gallon PVC lined tank which is located outside of the FMO building. This tank will receive the treated product from the primary nitrate waste treatment system and the AW waste from the solvent extraction system. The liquid in the tank will be kept in constant circulation by a pump. The circulation stream will be monitored for density and if the density of the solution increases above a set point equivalent to a maximum uranium concentration of grams per t

liter, an alarm will be sounded in the control room and the double block and bleed valves on the inlet streams will automatically close. (Double block and bleed valve operation is illustrated in Figure 2.3-4.) A controlled

stream will go to the Waste Treatment Facility for '

i further treatment.

2.3.2.2 ' Secondary Nitrate Waste Treatment 2.3.2.2.1 Reaction Tank A controlled stream of nitrate waste from tank V-103 will be received into the reaction tank located at the Waste Treatment Facility. This stream will be mixed i

with a lime slurry to adjust the pH to This t will' precipitate the metal ions and uranium that is in the aqueous waste liquid from solvent extraction.

The treated solution, a dilute slurry, will flow by gravity into the existing settling tank V-104.

( 2.3.2.2.2 settling Tank - Existing V-104 l

l 'The dilute slurry from the reaction tank will flow into the draft tube of the settling tank. A t may be added to enhance the settling of the solids. The

, clear effluent will be drawn off the top of the liquid

, in the tank and discharged to the nitrate lagoons. The l

Reference:

SNM-1097 Page 2.3-10 N. - _ . -

June 1, 1984 i

FIGURE 2.3-4 SLOCK & BLEED VALVE SYSTEM N/ N/ ' F LOW A A '

1 2 The system is controlled by a sequence controller that is activated by an OPEN or CLOSE signal. ;

NORMAL FLOW POSITIONS Block Valve #1 is OPEN Block Valve #2 is OPEN Bleed Valve #3 is CLOSED

! VALVE OPERATION CLOSE:

l To stop the flow, the CLOSE signal is activated. The valves operate in sequence. That is, when block valve

1 closes, then block valve #2 closes, then bleed valve

3 openc. If either block valve #1 or 2 should leak, the Jiquid will flow out through the bleed valve #3 onto the floor.

OPEN:

To open the valves,,to start the flow, the OPEN signal is activated. The valves operate in sequence. That is,

! when bleed valve #3 closes, then block valve #2 opens, then block valve #1 opens.

(Duplicate of Figure 2.1-8) l l

Reference:

SNM-1097 Page 2.3-11

4 June 1, 1984 i concentrated sludge will be removed out of the bottom tank outlet for filtering. '

The concentration of uranium and solids in this system L will be as follows:

l o The feed to the reaction tank should have a density of grams /cc, with a uranium concentration of t about ppm. '

o The lime treated solution from the reactor tank

, will have a density of grams /cc, with about t

] ppm uranium and about solids.

o The clear effluent from the settling tank should have a density of grams /cc with about ppm t uranium.

o The concentrated sludge removed from the bottom of ;

the settling tank will have density of about '

t grams /ml with about ppm uranium and about t

] 4 solids. t j 2.3.2.2.3 Filtration l ,

The sludge from the bottom of the settling tank will be I pumped into the secondary nitrate waste sludge filter

{ for solids removal. The filtrate will be returned to i

the tank and the solids will be discharged into five gallon pails. The filter cake collected in the five gallon pails will have about ppm uranium and t about 4 solids. t 2.3.3 Criticality Safety considerations i

The nitrate waste system is' designed to process nitrate i waste liquids containing low concentrations of uranium, l as noted in Section 2.3.2. Prior to the implementation i of UPMP, the uranium concentration in this nitrate waste 1 stream has been low, ppm or less. The sludge that t is removed from the-nitrate waste stream prior to )

discharge to tank V-103, has a uranium concentration of ppm. The sludge that is removed t

4 t

'*9' *~

Reference SNM-1097

June 1, 1984 i

from tank V-104 at the waste treatment facility has a uranium concentration of ppm. These t sludges are collected in five gallon pails, dried and stored for off-site processing. The uranium concentration in the nitrate waste system liquids are a factor of or more below the minimum critical t i

concentration for a homogeneous mixture of pure UO 2 and i water. As indicated by the results in Section 4.1.9.1

, the Kinf = 1.0 value is approximately 200,000 ppm U or equivalent to 250 grams uranium / liter.

The implementation of the UPMP nitrate waste system

between nitrate waste quarantine tanks and the discharge to the existing waste treatment settling tank V-104 will impact the nitrate waste system concentration control as shown in Table 2.3-1. The major impact will be the reduction of uranium concentration in the liquid discharge to the nitrate lagoon by a factor of t or more. . ; t i

The following sections describe the specific controls :

implemented in the UPMP nitrate waste treatment system to assure criticality safety.. These controls include not only those existing prior to UPMP, but also the process considerations which limit achievable uranium

concentrations. Also, extensive use.of geometry in the UPMP nitrate waste treatment equipment assures

! criticality safety. A final Section 2.3.3.4 is included

_ which discIsses the criticality safety of the nitrate waste system under normal and accident conditions.

r i 2.3.3.1 Uranium Concentration control in UPMP Nitrate Waste Collecting Vessel l Af ter the implementation of UPMP Nitrate waste treatment system, the operation of the FMO nitrate waste quarantine tanks will not change except that they will l discharge to the nitrate waste collecting vessel instead i

of the storage tank. The nitrate waste quarantine tank dump limits will not change prior to UPMP implementation and the uranium concentration is verified by a SPEC 20 i measurement to,be below ppm. This insures that t the uranium sludge is removed before the nitrate waste

.is transferred to UPMP-collection' vessel.

Page 2.3-13

Reference:

SNM-1097 l

l

June 1, 1984 l

TABLE 2.3-1 NITRATE WASTE URANIUM CONCENTRATION CONTROL Uranium Concentration Range (ppm U)

Prior to After Overall System UPMP UPMP Effect

~

Discharge from Unchanged t Nitrate Waste

-Quarantine Tanks Sludge from Unchanged i Nitrate t Centrifuge Nitrate Waste ---

Surge t Collecting Vessel Capacity *;,

I Nitrate Waste ---

Pre- t Vessel Treatment

t Filters ---

Rocovery* t

'1

'l Centrifuge Sludge ---

Recovery

t t

Nitrate Waste ---

Improved Product Aqueous Waste to ---

None t Existing V-103 Transfer to Improved t Existing V-104 Effluent to Improved t Nitrate Lagoon Filter Cake from .

Improved t V-104 t Hitrate Lagoon Unchanged t Discharge

Provides direct recovery and internal recycle of the uranium.

Reference:

SNM-1097

June 1, 1984

, The operation of the nitrate waste collecting vessel is detailed in Section 2.3.2.

i In addition to this, the nitrate waste collecting vessel i will be instrumented with density monitors interfaced with the UPMP process control system. If these density monitors detect values higher than the equivalent of t grams uranium / liter, the process control system will i automatically close all process inlet valves and sound an alarm to notify the control room operator of the condition.

2.3.3.2 Nitrate Waste Treatment Chemistry and Process Control 2.3.3.2.1 Chemistry i

The nitrate waste accumulated in the collection vessel will contain soluble uranium, carbonates, fluorides and

, numerous metal ions. A controlled stream of nitrate . ;

waste will flow from the collecting vessel to '

the tank. The liquid in the tank will be thoroughly mixed with an agitator. A t lime slurry will be t added to the tank to adjust the pH to about t

' This lime will precipitate the uranium, carbonates t and fluorides as calcium compounds and at the pH j of the metal ions will precipitate as hydroxides. t The pH of the nitrate waste stream is measured before i

' and after the tank. This controls the addition of 4 lime slurry to attain the proper pH. The t treated nitrate waste from the tank flows into

. -the vessel to stabilize the reaction. t l 2.3.3.2.2 Process Control i

^

The UPMP nitrate waste treatment system will be controlled by the UPMP process control' system. The flow rate through the process will be controlled by the

' liquid level in the collecting vessel. This level will be dependent on the liquid ' input from the various sources of nitrate-waste.

i -

l l

l Page 2.3-15

Reference:

SNM-1097

June 1, 1984 l

t t

e*

t t

t

t t

t The product in the tanks will be circulated and the discharge rate from the system will be controlled by the liquid level in the tanks. A small side stream will be diverted through a constant flowing sample loop to the i 1

UPMP control laboratory where a sample will be withdrawn from the sample loop and analyzed for uranium. The results will be transmitted to the UPMP process controller. If the uranium is less than ppm, the product will be continuously discharged to nitrate waste storage tank V-103. If the uranium is above ppm, an t alarm will be sounded and the product will be automatically diverted back to the collecting vessel for reprocessing.

i The density of the product will be monitored in the storage tank. If this should exceed a density which is equivalent to an upper limit of grams uranium per t liter, an alarm will be sounded and the product stream automatically diverted back to the collecting vessel for reprocessing.

i Page 2.3-16

Reference:

SNM-1097

~

s s

June 1, 1984 i

The product from the tanks will be discharged into nitrate waste storage tank V-103 located outside the 1 west end of the building. The aqueous waste stream from the aqueous waste vessels, acid raffinate from solvent extraction, will also be discharged into V-103. The liquid in the tank will be recirculated for thorough mixing and the level will be maintained by diverting a controlled side stream to the secondary nitrate waste tank located at the Waste Treatment Facility.

This stream will be monitored for density and, if it exceeds a set point equivalent to grams of uranium t per liter, the alarm will be sounded and the inlet

, valves will be closed. -

A proportional sampler will be located in the pipe line i between the product tanks and storage tank. The sampler will measure the volume discharged and obtain a i

. representative sample of the liquid for uranium analysis by the Chemet laboratory. ;

I '

The nitrate waste stream flowing into secondary nitrate waste reaction tank will be mixed with a controlled volume of 4 lime slurry to adjust the pH to t approximately This treated stream will flow by t gravity into the settling tank V-104. The pH of the inlet and discharge streams of the tank will be L monitored by the control system to adjust the lime

' slurry flow to maintain a pH. t

! The lime in the tank will precipitate the uranium and the metal ions that were added to the nitrate waste j stream by the aqueous waste. The solid will settle to

the cone bottom of the settling tank V-104 and the clear f liquid will be decanted from the top.

I 2.3.3.3 Geometry Control of Nitrate Waste Treatment System The UPMP' nitrate waste treatment system.will be i . implemented with extensive use of geometry control in equipment design, summarized as follows:

l l (1)

geometrically safe cylindrical t I

. tanks i

e

Reference:

SNM-1097 Page 2.3-17

June 1, 1984 (2) geometrically safa vessels (3) geometrically unaafe tanks t (4) geometrically safe filters t (5) geometrically safe cylindrical centrifuge t (6) ,

geometrically safe filter t (7) Geometrically safe pumps (8) geometrically safe diaphragm pumps (9) One (1) critically safe floor basin for liquid spills.

(10) Associated geometrically safe process piping.

(11) Material handling equipment in compliance with '

geometrically' safe designs

, Table 2.3-2 contains a .?isting of UPMP nitrate waste equipment.

1 2.3.3.4 Normal and Accident Conditions in the Nitrate Waste system As noted previously, the maximum uranium density normally present in the UPMP nitrate waste system will be less than grams / liter. -From Table 4-4 in t Chapter 4, the Kinf for a full density homogeneous mixture of UO 2 and H2 O with a U density no greater than

this will be in the range of 0.30-0.50. It can i

therefore be inferred that under normal conditions, the l

Kinf for any piece of equipment in the nitrate waste system will be well below a value of 0.90.

In the evaluation of nitrate waste system accident conditions, three significant conditions have been identified. These are: '

i a

i !

1 1

Reference:

SNM-1097 Page 2.3-18 -

i

O-June 1, 1984 TABLE 2.3-1 UPMP NITRATE WASTE EQUli? MENT LIST Normally Equiement contains Criticality Safety Lab- Equipment oetignation Uranium Geometry Description Description Centrifuge Yes Geometrically Safe cylinder t

' t Filters Yes Geometrically Safe t t

t Pilter Yes Geometrically Safe Slab t t

t Transfer Pumps Yes Geometrically Safe Slab & t Volume t t

Pumps Yes Geometrically Safe Volume t t

t Pump Yes Geometrically Safe Slab & t Volume t

~ t Pumps Yes Geometrically Safe Slab & , t Volume t

t Pump Yes Geometrically Safe Slab & t Volume t t

Transfer Pump Yes Geometrically Safe Slab & t Volume t t

Pumps No, Geometrically Safe Slab & t ppm U Volume t t

Pump Mo, Geometrically Safe Slab & t ppm U Volume t t

Pump Yes Geometrically Safe t

, t Pump Yes t Geometrically Safe slab & t volume t Pump No t Geometrically Safe Slab & t Volume t Tank Yes t Geometrically Safe Cylinder t t

Tank Yes t Geometrically Safe cylinder t t

Seal Pots No t Geometrically Safe Cylinder t Sump No t Geometrically Safe Cylinder t t

Tank Yes- t Geometrically Safe cylinder t t

Tank Yes t Geometrically Safe Cylinder t t

Tank Mo, t ppa U Geometrically Safe cylinder t t

1 j

l Page 2.3-19 i

Reference:

SNM-1097 f

,, er -- --- .w-- -- e-

f June 1, 1984 i

TABLE 2.3-2 UPMP :.ITRATE WASTE EQUIPMENT LIST (CONTINUED)

Normally -

Ccitains criticality Safety Egalpment Geometry Description Description Label Equipmen,t Lesignation Uranium Tank' Yes Geometrically Unsafe t !

t Yes Geometrically Unsafe t f Tank f 6 Yes Geometrically Safe Cylinder t Tank l

I No Geometrically Unsafe .

Tank t t ,

No Geometrically Unsafe t Tank t t

Yes Geometrically Safe with t Tank Neutron Absorber Panels t f

i i

, t ,

Geometrically Safe with t ,

Tank Yes Neutron Absorbe Panels t t

Process Piping Yes 1*,2*,3*,4",6* Pipe Geometrically Safe Cylinders and Slabs Floor Basin No 3.5". Thick Slab Critially Safe Slab with )

i Concrete and 1" Water j i Reflection i

Five Callon Pails Yes 11-1/4" 10, 13-1/2* Migh, 22 Geometrically Safe cylinder Liters t

P e

5

Reference:

SNM-1097

---. ..,., _y ,_. - .

t June 1, 1984

! 2.3.3.4.1 Failure of Concentration Control in Nitrate Waste System

[

As noted in Table 2.3-2 almost all process equipment in the UPMP nitrate waste system will be geometrically safe

and failure of nitrate waste system density control will i not result in a critically unsafe system. The only

! exceptions to this are the existing V-103 nitrate waste storage tank and the lime storage tanks. The lime t storage tanks are large geometric process support vessels not designed to contain uranium and are i discussed further in Section 2.3.3.4.3. Failure of concentration control in V-103 requires the following

. independent and coincident conditions:

(1) Failure of concentration control in the nitrate '

waste quarantine-tank system including not only

, failure to restrict high level uranium streams from

! the system, but also failure to detect the high U. ;

l- content via the SPEC 20 monitor. '

I (2) Failure of the UPMP (geometrically safe) nitrate

, waste treatment process, especially the j precipitation and filtration steps, to remove the i high level of uranium in the stream.

, (3) Failure of the UPMP process control system to datect the increase of the turbidity in the output fro.n

! __ the filters, failure of the density monitor on storage tank V-103, and failure of the process i laboratory to detect the high uranium concentration, l and the failure of the pH control system.

i For an unsafe condition to arise, all three of these independent cases must occur at the same time. It is therefore concluded that the criticality safety of tank i

V-103 is fully in compliance with the double contingency policy.

References SNM-1097 Page 2.3-21 l

June 1, 1984 2.3.3.4.2 Failure of Geometry Control in the Nitrate Waste System Failure of geometry control in the nitrate waste system will not result in a critically unsafe condition without the coincident and unrelated failure of the nitrate waste concentration controls.

2.3.3.4.3 Backflow of Uranium into Unsafe Geometry Vessels Backflow of uranium into the lime slurry tanks from the nitrate waste system will be prevented by use of a positive air break in the process piping from the lime slurry tanks to the nitrate waste system. This will be '

achieved by piping the lime slurry solution into the top of nitrate waste tank which will be vented to the atmosphere at a point above the solution level in the tank.

If this air break backflow prevention fails, however, '

the lime slurry tanks will still be critically safe

~

because of the nitrate waste system concentration control.

I

't Page 2.3-22

Reference:

SNM-1097

~

June 1, 1984 2.4 SCRAP PROCESSING 2.4.1 Current Scrap Processing Operation High grade uranium bearing scrap generated during fuel manuf acturing is currently oxidized in the REDCAP furnace and processed through the Uranium Purification System (UPS) digesting and conversion facilities located

( in the FMO building.

Low grade uranium bearing scrap generated by the waste treatment operation and the decon facility, and high grade gadolinia bearing uranium scrap generated in the Gad shop are placed in five gallon pails and shipped off-site for vendor processing. The purified uranyl nitrate is returned via tank truck and unloaded into the

~

FMO-X slab storage tanks for subsequent UPS conversion.

The new scrap processing facility will replace the ;

existing REDCAP furn' ace and UPS digest facilities as well as the off-site processing requirements.

i 2.4.2 UPMP Scrap Processing Uranium-bearing scrap will be processed for uranium recovery, purification and recycle to the fuel conversion process. Scrap processing unit operations include:

e oxidation and drying of sludge and scrap,

-

High grade scrap dissolving and dissolver solution filtration, l

e Low grade scrap leaching and leacher solution j filtration, i

e -Feed adjustment, e Solvent extraction recovery and purification, e Uranyl nitrate product concentration and storage, l

Solvent treatment.

i

[

High grade scrap dissolution will be accomplished in scrap dissolvers designed for a maximum of kg UO t equivalent per batch. Lowgradescrapwillbedissolve$

in safe geometry pipe dissolvers. Release of both high grade and low grade scrap materials to dissolution will be based on enrichment lots in order to minimize downgrading of enrichments and yet meeting production requirements.

i i

l Page 2.4-1

Reference:

SNM-1097

June 1, 1984 1

2.4.2.1 Scrap Processing Material Handling - Scrap Input i 2.4.2.1.1 Input Materials to the scrap processing system are generated from multiple sources as follows: ,

l 2.4.2.1.1.1 Wet Sludges I e Wet sludges from Fuel Manufacturing Operation (FMO) conversion. area

- Rad waste sludge

- Fluoride (ADU) sludge !

- Nitrate waste sludge l

- Wet powder '

e Wet sludges from the decon facility )

e Wet sludges internally generated in the UPMP area (Gd-contaminated): -

i,

- Rad waste lime sludge l

- Nitrate waste lime sludge i Wet sludges will be packaged in five gallon plastic pails having a polyethylene bag as an inner liner and 1 closed. The pail lid will be sealed. The maximum I

. weight of a five gallon pail.of wet sludge will not exceed 35 kg. All plastic pails generated in the FMO conversion area will be recycled to that area. All other pails will be recycled only within the UPMP area.

2.4.2.1.1.2 Non-Gd Powder and Pellets e . Dirty powder generated in the FMO conversion and fabrication areas e Non-recyclable' pellet hard and green scrap dispositioned from an outside storage area or the fabrication area e Dry filter knock-out powder from the decon facility The pellet scrap and dry filter knock-out powder will be packaged in polyethylene bags closed and placed in three and five gallon metal pails, respectively. The maximum weight per pail may not exceed 35 kg for either type.

Reference:

SNM-1097 Page 2.4-2 i

I

June 1, 1984 2.4.2.1.1.3 Gd-Bearing Scrap e . Dirty Gd powder e Non-recyclable Gd green pellet scrap e Non-recyclable Gd hard pellet scrap e Incinerator ash e Incinerated HEPA filter media The pellet scrap and Gd-bearing sludges are packaged in three and five gallon metal pails, respectively. The maximum weight per pail may not exceed 35 kg.

2.4.2.1.2 Output Solid Materials from the scrap processing facility will be:

e Insoluble filter cake containing very low levels of uranium which, after being washed, will be assayed, accountability weighed, tampersafe sealed and packaged in three or five gallon pails and placed in; 4'x4'x3-1/2' metal boxes for off-site burial.

e Non-combustible solids such as worn-out equipment, furnace muffles, piping, etc., which will be cut into pieces small enough to fit inside a 4'x4'x3-1/2' metal box, scanned and then transported to the decon facility for resorting and repackaging prior to final disposition.

-_-. . _ e Combustible waste packaged inside 4'x4'x3-1/2' wooden or cardboard boxes, scanned and transferred to the decon facility for re-sorting and repackaging prior to incineration.

2.4.2.2 Scrap Processing Material Handling - Process Considerations The following constraints have been applied in the design of the scrap processing facility:

)

i 2.4.2.2.1 With the exception of. incinerator ash, all sludges (both wet and dry) will be oxidized in the muffle furnace prior to dissolution. Oxidation is essential to remove all organics (oils, greases, binders), volatile i

i 4

Page 2.4-3 !

Reference:

SNM-1097 e-- ..,.-.c. ,y,_,w-,,.--- . - - , , , - - , - . , , . , , . - , , ,_ - --.-.-.--,,,----_r- ~--n ,,.,w~.,..,v,,- - -

,,-<s- ,,,

June 1, 1984 salts (NHgF, NH 3, HNO 3, SiF g, Nacl) and all combustibles such as wood, paper, and plastics. The wet sludges are assumed to contain 50 weight percent water (nominal) .

2.4.2.2.2' All scrap material and product movements into and out of this facility will be entered on the . computerized Manufacturing Information Control System (MICS) and a traveler (MICS) card will be attached to each container

, of material.

- 2.4.2.2.3 High grade scrap powder and sludges will be processed in three or five gallon pails. High grade pellet hard and green scrap will be processed in three gallon

. pails. -

2.4.2.2.4 All oxidized materials will be processed through a size reduction step, homogeneously blended within each pail, isotopically scanned for percent U 235, accountability weighed, tampersafe sealed, entered on MICS, and placed; inside a sealed polyethylene bag. Interim queuing of '

^ all pails of oxidized material will be provided by existing outside storage pads.

2.4.2.2.5 The production control function will retrieve oxidized materials from the pads for dissolution and recovery by 3 specified enrichment band in a minimum lot size of kg UO2 equivalent. t 2.4.2.3 Scrap Processing Material Handling - Sludge Transfer The process feed for scrap processing will be routed in and out of the UPMP building through the following areas e Sludge oxidation furnace input queue conveyors e -Oxidation furnace discharge queue conveyor e Gd scrap queue conveyor e Dissolver feed queue for oxidized sludge and incinerator. ash r

a Page 2.4-4 j

Reference:

SNM-1097

June 1, 1984 2.4.2.3.1 Sludge Queueing Pellet scrap in three gallon pails and sludge in five gallon pails (each with a maximum weight of 35 kg/ pail),

generated in FMO conversion and fabrication, will be transported to the FMO-X corridor queue. conveyor outside the oxidation area. The operator delivering the sludge will make a MICS entry transaction for each pail of sludge at the MICS terminal located in the corridor.

The operator will place a spacer skid on the storage conveyor and then place a pail on each spacer skid. Wet sludge will be packaged in five gallon plastic pails lined with a plastic bag sealed with a twist lock. Dry

. powder and pellet scrap will be packaged in the same manner except the external container will be a three or five gallon pail.

The pail conveyors are located 14" above the floor and can each store up to 30 three or five gallon pails of, ;

sludge (on spacer skids). As required by the process '

control plan, the operator will manually transfer cans in batch quantities from the storage conveyor (located on the south wall of the corridor) to spacer skids on the oxidation furnace queue conveyor located on the north side of the corridor. A single pail of scrap or sludge will be processed in the oxidation station at a time. When a pail is needed, the operator at the enclosure will activate the powered conveyor and transfer it, still on its skid, to the bottom entrance door of the enclosed bucket lift. (See Figure 2.4-1.)

The pail and skid will be transferred automatically via the pail lift to the siphon station within the enclosure.

2.4.2.3.2 Boat Loading When a pail of scrap or sludge arrives at the siphon station, the operstor, working through openings in the enclosure, will remove the pail lid and set it aside.

If the material is wet sludge, the twist lock is removed from the plastic bag. If there is free standing liquid over the sludge, the operator will open the plastic bag and use the siphon wand to siphon off the excess liquid.

(The removal of the siphon wand from its hanger will automatically activate a pump and associated valving.

l l

Page 2.4-5

Reference:

SNM-1097

June 1, 1984 FIGURE 2.4-1 t

t t

d'

,e t

t t

i

t t

t t

}

t t

Page 2.4-6 Reference SNM-1097

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

June 1, 1984

+

Replacing the wand on the hanger will stop the pump.) ,

The operator will then seal the plastic bag with the '

, same plastic twist lock and move the pail of scrap or sludge to the skid stop where the pail is mechanically

': lifted from the spacer skid and pushed to the pail dump station. At the pail dump station, the pail will_be tilted and the operator, working through openings in the i- enclosure, will pull the plastic bag with sludge or dump l the pail contents, if scrap , into the inclined chute and slide it into the oxidation boat below. A ram pushes i downward on the plastic bag both piercing it and 3 assurino no pile up of material in the center of the t

deep rectangular boat. Only the contents of a t

single pail will be dumped into each~ oxidation boat. i i

The operator will insert a card into the MICS terminal ;

and activate the " ENTER MICS" button to record the transfer of material to this station.

I

-2.4.2.3.3 Pail Rinse Operation '

The operator will manually move the empty pail to the rinse station. At the rinse station, the operator, t ' working through' openings in the enclosure, will use the t

DI water rinse wand to clean the interior of the bucket and the siphon wand to remove the rinse water. The rinse water plus decant liquid will be transferred i directly to the rad waste accumulator tank. The lid to i the pail will be clamped on and the operator will then place the empty pail with the lid back onto its spacer skid. The pail will be returned to the conveyor via the bucket lift after which the operator will dispatch the next incoming bucket of sludge.

2.4.2.3.4 Gadolinium (Gd) Fabrication and UPMP Generated Sludge Handling l Gd-bearing sludges generated within the UPMP facility and the Gd fabrication area will be queued on skids on

the conveyor feeding the Gd dump station. (See Figure 2.4-2.) The operator will manually transfer one three j- or five gallon. pail of sludge or pellet scrap at a time from the cart to the conveyor. Receipt of each pail of

sludge at this station will be accountability weighed

and an entry made on a MICS terminal.

I i

Reference:

SNM-1097 Page 2.4-7 s

-n ,n-,w --,-.n. , - , - - . . ~ . ..,n.--...,,,,,,-.,,a- r ,.._,n,--,--,nn.---n.n --~~.---,n-,,--,,_.

i June 1, 1984 l

The sludges generated in the UPMP rad waste and nitrate waste systems will be collected at the centrifuge hoppers in three or five~ gallon pails, weighed, and the

! data entered into the MICS system. The pails will then

be placed on a cart and transferred to the second pail l- dump station queue conveyor.

The transfer of the sludges at the Gd pail dump station will be the came as that discussed in paragraph 2.4.2.3.2.

2.4.2.3.5 Oxidative Furnace Stoking and Boat Dumping i

The oxidizing furnace is a gas tight muffle type furnace '

which has an Inconel, "D" shaped section with guide rods on the inside to guide the boats through the furnace and prevent boat jamming. The maximum furnace operating 1 temperature will be C( F) and the t i maximum furnace input production capacity will be -

3, l' kg/ hour of wet ( t water) scrap. Each t furnace boat will hold one 35 kg (gross) pail of scrap.

The boats will pass through the furnace in a string pushed by the timed pusher mechanism. The nominal rate l is feet per hour. (The furnace muffle is sized for t boats, each inches high by inches ',

wide by inches long.) I i

1 A swinging overhead hinged door at either end of the muffle will allow one boat to be pushed in and one out when the pusher is activated. The furnace ends are

enclosed and vented to the exhaust scrubbing system.

The muffle extends out of the furnace and has a t

, long cooling jacket section on both ends to protect the enclosure and to cool a boat of oxidized sludge.

' Air flow through the furnace will be in the laminar to slightly turbulent range and controllable. The furnace offgas scrubber will use water as the scrubbing medium and is sized for a stu/hr heat load at t scfm flow rate. The offgas from the scrubber is t saturated with water and will go to the vessel offgas system. The scrub water will go to the nitrate waste accumulation tanks.

i

Page 2.4-8 l Reference SNM-1097 t

~

June 1, 1984 FIGURE 2.4-2 t

t t

. ; t t

t t

t Page 2.4-9 References SNM-1097

June <1, 1984

~

The operator will start the sequence to feed a loaded boat in the furnace by activating the furnace " STOKE" button. (The stoking ram will not activate until the preset stoking time cycle has elasped.)

2.4.2.3.6 Oxidation Furnace Discharge Operations At the end of the furnace, oxidized sludge will be dumped from each boat and will fall by gravity into a inch thick by inch hopper t feeding a roll crusher below. Pneumatic vibrators installed on the hopper feeding the roll crusher will minimize powder buildup.

. I The crushed oxide will flow through a inch t diameter rotary feed valve via a inch t diameter transition chute into a three or five gallon metal pail. A load cell will be used below each pail to

, limit the maximum gross weight of the pail to 35 kg. I(

the weight limit is reached, a wafer valve will automatically close and halt powder flow. As in the case of the oxidation feed operation, oxidized pellet hard or green scrap will be restricted to three gallon pails. The feed hopper is vented 'through a pre-filter unit and a HEPA filter housing to maintain dust control 3 in the enclosure. '

An empty pail cannot be introduced to the can fill

.- station until the pail just loaded is conveyed to the blend station. (See Figure'2.4-3) At the blend station, the operator will physically load and clamp the pail to the blender mechanism. The clamp on top'of the blender is designed such that it facilitates homogenous blending. After blending, the pails will be conveyed through an airlock to an isotopic scanning station.

After isotopic scanning, the pail will be accurately weighed on a suspended load cell, tampersafe sealed and then slipped inside a polyethylene bag. The tampersafe seal number and can weight will be -entered on MICS and a card attached to the pail. The pail is then conveyed to the queueing conveyors inside the pail / waste box handling station. At this station, the pail will be accountability weighed and a second MICS entry made.

Page 2.4-10 References SNM-1097

June 1, 1984 FIGURE 2.4-3 SCRAP PROCESSING PROCESS FLOW t

t t

, t t

t t

. ; t t

t t

t Page 2.4-11 References SNM-1097

O June 1, 1984 2.4.2.3.7 Handling, Staging and Transfer After the second MICS entry is made, the pails will be placed on the pail holding conveyor to await transport to the storage pads. Three and five gallon pails of oxidized sludge or scrap will be routed to and from the existing outside storage pads via a forklift truck. All pail movements to and from the area will be entered on a MICS terminal.

A pallet with pail separators will be used for the safe transport to and from the pads. Incoming pails of oxidized sludge released for dissolution will be routed to specific dissolver conveyor queues upstairs. I 2.4.2.4 Scrap Dissolver Operations The dissolver input dump hoods and batch queueing conveyors are located on the northwest mezzanine floor ;

level of the facility. (Figure 2.4-4).

A schematic representation of the entire scrap processing flow is shown in Figure 2.4-5, page one, with product concentration shown on page two of the figure.

1 2.4.2.4.1 Queueing i Pails of scrap material, as released by the production control function in > kg 00, equivalent lots, and t will be routed to the mezzanine floor queue area via a pail lift from' the pail entry area below. The operator will select the conveyor queue and will route each pail of scrap to the appropriate queuing conveyor feeding each dissolver or leacher. Each of the charging hoods has an automated pail dump mechanism. The dissolving / leaching operation will be done in three unit types e high grade scrap dissolvers t e low grade scrap leachers t e low grade leacher t Each pail will be moved into the pail dump enclosure and authorized for dumping at that station. The tampersafe Page 2.4-12

Reference:

SNM-1097

June 1, 1934 FIGURE 2.4-4 PROPOSED EQUIPMENT LAYOUT - SECOND FLOOR (ot(,,d, Dissolver queues E XTR A ctich

!!c'?Rc.nil'!ail*a m.= coman e.m.

~: n (g [( yjy ;gra=,=a=- i, (9 [ c^a '*'

.steseeMyst ecauseen 7

i # ,, ,

Y ,s m

.4. =c.

..=eine.me .

NITRATE & RAD WASTE- '

00000' 7

owaten Page 2.4-13

Reference:

SNM-1097

June 1, 1984 FIGURE 2-4.5 SCRAP PROCESSING PROCESS FLOW t

t t

t J

3 t

t t

t

. ; t t

t t

t 3

t t

., t t

t t

Page 2,4-14 References SNM-1097

l l

June 1, 1984 l

FIGURE 2-4.5 l SCRAP PROCESSING FROCESS PLOW l

(CONTINUED) t t

t

, t i t t

t t

t

. ; t 9

, t l t t

l t i t t

t t

i i

! t t t

, t

!. t i' t t

t i t t

t t

. t t

t t

Page 2.4-15

Reference:

SNM-1097

June 1, 1984 seal will be removed and the seal number entered on MICS.

The dump rate to each dissolver will be dependent upon the dissolver feed rate as specified in the Production Requirement Operations Document (PROD) and will vary according to the type of scrap being processed and the unit type being used. The high grade scrap will be dumped to the feed hopper. Pulsed air actuated bridge brekkers will be used to prevent bridging and holdup of oxidized scrap.

2.4.2.4.2 Dissolution 2.4.2.4.2.1 High Grade Scrap Dissolving / Filtration High grade scrap will be processed in kg 00 2 i equivalent batches per dissolver. Each batch dissolver i

cycle includes the following steps: -

i, e Fill / heat-up e Dissolving i e Empty / filtration / rinse During the dissolving sequence, nitric acid and aluminum nitrate will first be charged to the dissolver. The l I liquid in the dissolver will be heated by steam injection and recirculated until the proper reaction l

I temperature is reached. The control system will then :

permit operation of the dissolver feed system. The scrap will be fed to the dissolver from the feed hopper at a uniform rate via inch rotary feed valves. t 3 flow indicating device connected to the process control system will be used to assure that feed is flowing to the dissolver. The dissolver will be operated at a negative pressure with respect to the dissolver hood.

During the dissolving cycle, dilute fluoride waste i treatment product or DI water will be sprayed into the dissolver to minimize particulate entrainment and solids accumulation on the wall of the dissolver. The dissolver solution will be circulated through a heat Page 2.4-16 Reference SNM-1097 ,

June 1, 1984 exchanger to remove the heat of reaction. Moist air and oxides of nitrogen will exit the dissolver via a t inch cylindrical de-entrainment pad and a inch t diameter reflux condenser. The resulting condensate will be returned to the dissolver. The non-condensable gases from both dissolvers will be combined in the dissolver off-gas header and routed to the NOx absorber.

The prdcess control system will monitor and control the number of pails entering the dissolver and will compare this with the preset pail count for the batch. After the last pail.of scrap is charged to the hopper, the process control system will prevent further dumping.

The process control system will set a delay timer to assure complete dissolution. Digestion will be continued until the density of the dissolver solution stabilizes. At the end of the dissolving cycle, the

, dissolver solution will be at approximately gm U/1 t with a specific gravity of approximately '

t At the end of digestion, the dissolver solution will be pumped from the dissolver through the inch diameter t filter, through the inch diameter product t cooler, and into the dissolygr surge tank. (The filtration rate will change as the solids loading changes the flow resistance of the filter cake) . If the solids loading is high, increasing the pressure drop across the filter and decreasing the filtration rate, the flow will be automatically directed by the process control system to the alternate filter.

When the dissolver is empty, flush soluti n will be added via the dissolver spray and pumped through the system into the dissolver surge tank. The flush volume will be approximately filter hold-up volumes. When t the rinse sequence is complete, the filter (s) will be back-pulsed with air motivated filtrate to dislodge the filter cake, and the slurry subsequently back-flushed to one of the pipe leachers for a second digest cycle.

2.4.2.4.2.2 Low Grade Scrap Processing The low grade scrap material is characterized by a relatively low uranium content and a high weight percent of insoluble solids. This material will normally be i

Page 2.4-17 References SNM-1097

I June 1,1984 stored in three or five gallon pails and contain less than a safe batch of uranium.

Low grade scrap leaching is a batch controlled process t that is limited to one pail of dry feed per batch.

2.4.2.4.2.2.1 Leaching The leaching operation will be done in either of t l

'long, inch diameter leachers. The leachers t can be run simultaneously depending on processing needs.

j The feed to the leachers will be limited by the process

<1 control safe batch) system in eachto one five gallon leacher batch. pail of material Oxidized low gra (3e i scrap containing up to weight percent uranium will t t be batch charged to the leacher. I Prior to charging of the scrap material aluminum nitrate and dilution water will be fed to the leacher and the . ; i

!. its associated recirculation tanks. Circulation between' :

the tanks will then be started. The feed entering the i top of the leacher will form a wet slurry. After the l initial heat-up to a preset temperature, concentrated '

nitric acid addition will be started. The slurry will i be recirculated between the tanks as the digest process ; ;

goes to completion. i The leacher off-gas will exit the leacher via a inch t l diameter de-entrainment pad. The condensables (H 2O and HNO 3 vapor) will be recycled to the leacher and the non-

. condensable gasses will be combined with the dissolver off-gas and routed to the NOx absorber.

2.4.2.4.2.2.2 Leacher Solution Filtration When the dissolution is complete, the leacher slurry containing < grams of uranium per liter in solution t will be pumped by the leacher slurry pump from the ,

. leacher to a filter. The filtration rate through the f i'

filter is determined by the pump parameters and/or the solids loading or. the filter. The filter is sized such that with a nominal single batch the cake thickness will be thick. t J

J i ;

Page 2.4-18

Reference:

SNM-1097

June 1, 1984 Prior to filtered solids discharge, the filter will be air pressurized to blow all residual feed through the cake and out to the inch diameter filtrate t collection tank. The solids will be rinsed with either fluoride waste treatment product or DI water to displace interstitial liquor held up in the filter cake. The filtrate will be routed from the filter to the filtrate collection tank. A turbidity meter in the recirculation line on this tank detects bleed through and provides an input to the control system which determines whether the current filtrate stream is to be recycled to the leacher or pumped through the inch diameter leacher t product cooler to the leacher surge tank.

At the end of the rinse sequance, a final extended air blowdown will remove interstitial liquor in the cake.

The filter cake will be funneled into a five gallon pail. The insoluble solids will be weighed, sampled and a MICS transaction will be made for material - i accountability. Total uranium losses with this burial material (undissolved plus interstitial) are estimated to range from 4 to 4 of the uranium in the t leacher feed or typically grams uranium / pail, t 2.4.2.4.2.2.3 Residue Leacher An inch diameter utility leacher is provided for t processing miscellaneous scrap materials and/or "hard to handle" low grade scrap such as HEPA filter media, HEPA filter ash and insoluble solids rework. Due to process considerations, this material operation will be limited to < one safe batch and will be controlled by the process control system.

The leacher tank and its associated inch t diameter recirculation tank will be charged with a pre-determined amount of dilution water. A single pail containing no more than kg of material with a t variable uranium content will be charged directly into the in the leacher tank. Hot nitric acid will be Page 2.4-19

Reference:

SNM-1097

June 1,1984 metered into the tank and the reaction temperature monitored by the process control system.

If the reaction temperature exceeds a pre-set level, the acid addition will be terminated automatically. After acid addition is complete, the tank will be circulated for hours depending on the material +

type. At the completion of the digest step, the leach solution will be either pumped to the inch leacher t recirculation . tanks as dilution water or processed through the leacher filter. The action to be taken will depend on the original uranium content of the scrap being processed (e.g., very low uranium concentration scrap digest material will be used for leacher dilution '

water).

The leacher containing the insoluble solids will be

,' rinsed, allowed to drip dry and removed from tha vessel using the hoist. The basket will be positioned i, above the basket dump station. The pinned bottom plate of the basket will be dropped and the solids will be discharged directly to a five gallon pail. The pail will then be routed to solid waste handling for interim storage, assayed, and ultimately sent offsite for burial. i 1

2.4.2.5 Feed Adjustment The surge tanks for the dissolver and the leachers will contain the filtered products of the respective operations. The uranium concentration in the dissolver surge tanks will be f, gm U/1 and that of the t leacher will be 4, gm U/1. The target feed t concentration for the solvent extraction process is t gm U/1 to gm U/1. Each vessel is provided with t density measuring devices which are tied into the UPMP process control system.

The require 6 final adjustment of the feed solution composition for the solvent extraction process will be carried out in the feed adjustment vessels. t vessels are provided so that receipt of t solutions, sampling and adjustment of the uranium concentration and acidity can occur while simultaneously

~

References- SNM-1097 m

L -

June 1, 1984 holding or transferring adjusted feed to the solvent

. extraction feed metering vessel.

The solvent extraction process controls employed for uranium recovery and purification are designed for a range of uranium concentrations in the feed depending upon the blend of solutions being processed. Dissolver solution will be sufficiently high in uranium concentration that it may be blended with low grade solution. The anticipated range of uranium concentration in the solvent extraction feed will allow the processing of up to kg U/hr through a "

t extraction column with a resulting dilute product concentration up to ga U/1. t Preparation of solvent extraction feed will be accomplished by blending together various amounts of process solution, material for rework and reagents.

Process solutions are drawn from the dissolver and - i leacher surge tanks and the rework material from the '

UNH rework tank. The blended volumes depend upon the uranium concentration and enrichment in the various component streams and the desired concentration and enrichment for solvent extraction processing. Primary enrichment control is accomplished at the outside storage where scrap is segregated into enrichment bands based on measurements of~the material. Acid and D1 water may be required for adjustment of the acidity and uranium concentration, and aluminum nitrate for final adjustment of the A1/F ratio for corrosion control and/or to prevent formation of non-extractable uranium fluoride complexes. A sample stream from these feed adjustment tanks will be routed to the UPMP process laboratory where free fluoride will be monitored.

Uranium concentration and acidity will be obtained from inline instrumentation.

The adjusted feed material will be filtered through a final filter prior to solvent extraction to remove post-

' precipitated material that arises from solution cooling and aging.

2.4.2.6 Solvent Extraction solvent extraction technology will be employed to recover and purity uranium from the blended feed Page 2.4-21 References SNM-1097

l June 1, 1984 solution. Separation of the uranium from soluble !

contaminants in the solution will be accomplished by t preferential transfer of the uranyl nitrate to an ,

organic phase, scrubbing of the organic phase to remove trace contaminants, and stripping of the purified uranyl nitrate from the organic using slightly acidified, warm water.

The organic phase utilized in the process will be a mixture of tri-butyl phosphate and a normal paraffin hydrocarbon diluent (N-dodecane). The diluent affects the density and interfacial tension of the organic to l values that permit practical operation of the !

. solvent extraction columns. Extraction of the uranyl l nitrate by the organic results from the formation of l non-ionic complexes between uranyl nitrate and the I organic in a solar ratio of The transfer t !

equilibrium is strongly affected by the presence of high !

in the aqueous ;

i concentrations phase. of common ions The reverse transfer is fac(NO3)ilitated by the '

i absence of the common ions (NO"3) . Extraction of :

nitric acid and contaminants into the organic phase is a !

function of the organic phase solvent saturation.

The solvent extraction process w'ill employ standard i techniques for contacting the aqueous and organic 1 ,

phases. These columns consist of vertical. cylindrical l vessels (pipes) through which the aqueous and organic t phases are passed. The heavier, aqueous phase will be introduced at the top of each column and the lighter ,

organic phase at the bottom. Thorough mixing of the two phases will be accomplished within the columns.

The solvent extraction system employs extraction t columns, inch diameter columns and a t L inch diameter column, to perform the three operations of estraction/ scrubbing / stripping. The uranyl nitrate feed stream will be purped into the top of the column and pass downward. The ratio of organic to aqueous will be !

controlled to achieve proper uraninum saturation of the ;

i

+

l l

i Page 2.4-22 l

Reference:

SNM-1097 i

i i

June 1, 1984 organic ( gm U/1 physical limit) thus inhibiting t extraction of trace contaminants into the organic phase

! and simultaneously controlling uranium losses to a l target of less than gm U/1. t i The AW 1eaving the column will pass through a inch t diameter decanter for removal of entrained organic and will then flow to the inch diameter t aqueous waste recirculation tank.

A sample stream will be continuously pumped to the UPMP l laboratory where it will be periodically measured. This i '

process control measurement provides near real time knowledge of the loss to the aqueous waste stream.

The AW stream will overflow to either of t

! quarantine tanks where it will be recirculated and again sampled for uranium content by the UPMP laboratory. If the uranium content is greater than ppm (the -

i, t current dump limit to V-103), the batch volume will be recycled to the process for rework. If the aqueous waste stream is within the uranium target level, batch l disposal will take place through the gallon t existing nitrate waste tank (V-103) for further treatment at the secondary nitrate waste treatment facility.

The organic phase leaving the of the column will t pass through the inch diameter product surge tank to t become the feed (uranium-rich) stream for the next process stage.

l h

i 1

i Page 2.4-23 References 8NM-1097 L

4 June 1, 1984 The uranium laden solvent leaving the column will be l pumped from the inch diameter product surge tank to t the next process step where the uranium will be stripped l back into the aqueous phase. Dilute uranyl nitrate

product from the process column will be passed through -

a inch diameter product decanter to remove any. t

entrained organic, and then overflow to the inch t diameter product feed tank..

Geometrically-safe disengaging sections are provided at the ends of each solvent extraction column for entrained ;

phase separation. Each disengaging section will provide !

the necessary cross-sectional area to reduce the flow ,

velocity to a level at which phase separation will ' '

occur. Particulate impurities that enter the solvent extraction columns will tend to accumulate at the controlled aqueous / organic interf ace.

2.4.2.7 Product Concentration -

3, !

During start-up of the solvent extraction process, the I

dilute product will be recycled from the product !

decanter to the UNH rework tank until sample analysis by the UPMP laboratory system indicates an acceptable product quality (i.e., trace metal impurities). The , ,

! dilute product will then be routed to the inch ,I '

diameter product concentration feed tank or during !

enrichment changeover, the product will be collected in i __ __

an UNH dilute product surge tank. The dilute product will be pumped from either tank to the foot long inch diameter product concentrator reboiler. t

~ The overhead from the reboiler will enter the t i

, tall inch product concentrator. t i

1 l

I l !

I Page 2.4-24 l

{

Reference:

SNM-1097 r

l; June 1, 1984 The uranium concentration will be increased from < t ga U/1 in the feed to between ga U/1 in the t bottoms (target ga U/1).The maximum allowable t concentration is ga U/1. Overheads from the t concentrator (water vapor and trace RNO 3 ) will be condensed in the inch diameter product t concentrator condenser supplemented with DI water make-up, and recycled to the column as warm strip water from the strip water make-up tanks.

[ The concentrator product under density control will flow l by gravity through the product cooler into i

product accountability tanks. The full tanks will be-recirculated and sampled for product quality in the UPMP l process laboratory. Uranium concentration control will l be accomplished using in-line measurements t l A MICS transaction t will be made and the entire solution volume will be transferred to interim product slab storagg t tanks located in FMO-X. Storage tank selection will be '

based on enrichment. The tank will be flushed forward with an appropriate volume of nitric acid or DI water to purge the transfer line, and~ attain the end-point concentration of ga U/1 as nominal feed to the UNH t conversion process.

l 2.4.3 Criticality Safety Considerations i

! As in the case of the fluoride waste, rad waste and nitrate waste systems, criticality safety of the UFMP L scrap process operation is assured by a combination of l operational / process controls, chemistry limitations, I

uranium mass limits and geometry control. The use of l geometry control is especially emphasised because of the

! higher uranium concentrations and throughput rates l present in this system as compared to the process waste

! streams.

The nest section gives a brief summary of criticality l safety limits and parameters which are applicable to the

controls implemented 'in -the scrap processing operation.

l These are based on the in-Jepth discussions presented in Chapter 4. After this summary, a discussion is Page 2.4-25 Reference SNM-1097 L.

June 1, 1984 presented of.the application of these limits and parameters to the criticality safety considerations for

! the scrap processing operation. Following this, a final section is included which covers normal operations and accident conditions.

2.4.3.1 Criticality Safety Limits and Parameters Applicable to scrap Processing Operations As described in Section 2.4.2, the UPMP scrap processing operation consists of four basic systems: scrap ,

oxidation, scrap dissolving, solvent extraction, and uranyl nitrate product concentration and storage. The uranium processing equipment in these four systems is of -

eight typss:

Three and five gallon palls I

e Oxidation furnace boats e Dissolvers . ;

e vessels '

e Leachers e solvent extraction columns e Nitrate waste tank V-103 e Miscellaneous types of critically safe equipment (such as pumps, filters, seal pots, sumps, floor ;

basins, etc.) :

With the exception of the nitrate waste tank, equipment in each of these categories is critically safe primarily because of geometry and materials of construction and

+

secondarily because of mass, concentration or moderation control. Criticality safety of tank V-103 is based on concentration and process / chemistry controls. Before a more in-depth discussion, however, a summary of the criticality safety parameters applicable to this equipment has been tabulated based on the discussions in Chapter 4. These are presented in Tables 2.4-1 through 2.4-9.

Table 2.4-1 lists critically safe geometric dimensions for equipment containing mixtures of 002 and water which have simple geometries or small volumes. Tables 2.4-2 through 2.4-6 summarise key Monte Carlo keff calculations for three and five gallon pails, dissolvers, leachers, tanks and solvent extraction columns.

Reference SNM-1097

June 1, 1984 TABLE 2.4-1 GEOMETRICALLY SAFE PARAMETERS FOR 00 2 + H 2O MIXTURES A. Safe Geometry Parameters Homogeneous Heterogeneous Mixtures Mixtures safe cylindrical Diameter 9.5 Inches 8.3 Inches Safe Slab Thickness 4.45 Inches 3.6 Inches safe Volume 24 Liters 13 Liters 3

B. Geometrically Safe

Dimensions for Inch ( Inch OD) t Infinite cylinders with Homogeneous UO 2 + H2O Mixtures Wall Inner Pipe Thickness Diameter Wall Material Schedule (Inches) (Inches)

PVC 0.165 t Stainless Steel 304 0.250 t Carbon Steel 0.365 t

GEKENO Keff + 3a - bias 1 0.97 d

Page 2.4-27 References SNM-1097

June 1, 1984 TABLE 2.4-2 NEUTRON MULTIPLICATION FACTORS (Keffs)

FOR THREE & FIVE GALLON PAILS I

W/F Volume GEKENO

, Ratio Keff j; o A. Three Gallon Pails 1.0 with Heterogeneous 00 2 +HO2 2.0 3.0 t

Mixtures 4.0 t 5.0 t 6.0 t WF GEKENO H,0 Keff f; a B. Five Gallon Pails 0.10 +

with Homogeneous 0.1573 U0 2 +HO2 0.2183 t Mixtures 0.2711 t 0.3172 +

0.3486 0.35 t 0.40 t 0.45 t 0.50 t NOTE: WF = Weight Fraction W/F = Water to Fuel Page 2.4-28 References SNM-1097

June 1, 1984

. TABLE 2.4-3 Neutron Multiplication Factors (Keffs) for Dissolvers with Neutron Absorber Panels t (Heterogeneous 00 2 + H2O Mixtures) w/F Volume GEKINO Ratio Keff + ci 1.0 t 2.0 , x. t 3.0 . t 4.0 t 5.0 t 6.0 t 3

TABLE 2.4-4 Neutron Multiplication Factors (Keffs) For Cylindrical I'nch Schedule Pipe Tank Leachers

t (Heterogeneous U + H 2 O Mixtures)

\

W/F GEKENO Volume Ratio Keff + o 1.0 '

t 2.0 s t 3.0 t 4.0 t 5.0 ,

t 6.0 ~

t 8.0 t 10.0 t

Infinitely Tall 1 :.

E

+

, s

\ ^\

.: ) i

Page 2.4-29

Reference:

SNM-1097 v< U' '

s

....t

June 1, 1984 TABLE 2.~4-5 Neutron Multiplication Factors (Keffs) For Vessels with Heutron Absorber Panels t (Homogeneous UO 2 + H2 O Mixtures)

GEKENO WF HO 2 Keff + o 0.8972 (100 gm U/ liter) t 0.7745 (250 gm U/ liter) t 0.6787 (400 gm U/ liter) '

O.5202 (750 gm U/ liter) i 0.10 t 0.20 t 0.25 t 0.30 ; t 0.40

t 0.50 , t I

_ . . . . . _ . , TABLE 2.4-6 Neutron' Multiplication Factors (Keffs) For Solvent Extraction Columns (Homogeneous UO 2 + H2O Mixtures)

GEKENO WP H,0 Keff. + o O.10 t

! 0.20 t 0.25 t 0.30 t 0.40 t 0.50 t i

) ,

s Page 2.4-30

Reference:

SNM-1097-

June 1, 1984 TABLE 2.4-7 SAFE BATCH LIMITS Mass, kg U Uncontrolled Geometry 16 (18.1 kg UO2)

(Homogeneous UO 2 + H2 O Mixtures)

Uncontrolled Geometry 16 (18.1 kg 002)

(Heterogeneous UO 2 + H2 O Mixtures)

Vessels

215 t (Homogeneous UO 2 + H2 O Mixtures)

Dissolvers

400 t (Heterogeneous UO 2 + H2O Mixtures)

45% of minimum critical mass assuming failure of neutron . ;

absorber panels but integrity of

vessels . t TABLE 2-4.8 UO 2 + H 2O Kinf VALUES (Homogeneous Mixtures)~

. GEKENO WF H,0 Kinf + o 0.0 t 0.005 t 0.05 t 0.10 t 0.20 t 0.25 t 0.30 t 0.40 t 0.50 t 0.75 t 0.7745 (250 gms U/ liter) t 0.8972 (100 gm U/ liter) t Page 2.4-31

Reference:

SNM-1097

l l

June 1, 1984 TABLE 2.4-9

00 2 + H 2O Kinf VALUES 7 HETEROGENEOUS MIXTURES i

l W/F GEMER ~ i Volume Ratio Kinf + o 1.0 t t

2.0 t

.t 3.0 .

t t

4.0 t t

5.0 t t

.6.0 +;

~t 8.0 t

- 10.0 Page 2.4-32

Reference:

SNM-1097

4

. June 1, 1984 Safe batch limits are tabulated in-Table 2.4-7 both for uncontrolled geometries-and for t vessels and dissolvers with the assumption that the neutron absorber panels have failed.

Finally, Tables 2.4-8 and 2.4-9 contain Monte Carlo infinite neutron multiplication factors for homogeneous and heterogeneous UO 2 + H 2 O mixtures over a range which '

' includes optimum moderation. Table 2.4-8 also includes Kinfs-for dry (moderation control) and dilute (density control) mixtures. Table 2.4-9 for heterogeneous mixtures is included for comparison with the homogeneous Case.

2.4.3.2 Scrap Processing Operations Criticality Safety As noted in Section 2.4.3,.the use of geometry control I

in the UPMP scrap processing operation has been especially emphasized as a primary control because of- i '

uranium concentrations and throughput rates which are significantly higher than those in the three UPMP liquid waste streams. In conformance with the double

contingency policy, however, scrap processing equipment

criticality safety is also assured by secondary (and in some cases tertiary) controls such as geometry, uranium concentration, uranium mass, or moderation. These secondary (and tertiary) controls are usually the same type for individual pieces of equipment in each of the 4

~ ~ ~ ~ ~ eight basic categories defined in the previous section.

, Table 2.4-10 summarizes this, subject to the following considerations:

2.4.3.2.1 Three and five gallon pails are used for the handling and storage of high and low grade scrap and sludges.

4 Three gallon pails are used for " heterogeneous" pellets and hard ' scrap _and five gallon pails are used for powder and sludges (materials which form " homogeneous" mixtures with water). This. distinction is important since as-indicated in Table 2.4-2, the three-gallon pail is geometrically safe for heterogeneous U0 2 and water mixtures by virtue of its diameter and height whereas the five gallon pail is not. (The five gallon pail is geometrically safe for heterogeneous'U(4.0)O 2 and water mixtures and may be used in operations with uranium scrap with enrichments no greater than this). With

. geometry as the primary control', three and five gallon Page 2.4-33

Reference:

SNM-1097

_ _-_-.s_ _ . . . - _ _ _ _ . - . _ _ _ . _ . .- _ _ _ . ._ _ _ _ _ . - _ _ _ _ _ _ - _ _ ,

June 1, 1984 TABLE 2.4-10 SCRAP PROCESSING OPERATIONS CRITICALITY SAFETY

SUMMARY

Controls Typ* of Equipment Contents Framary Secondary other IA. Five Gallon Pall High Grade Scrap Powder Geometry Volume Mass / Moderation (Ref. Table 2.4-2) (Ref. Table 2.4 1) Control

~

Wet & Low Grade Scrap Geometry volume Safe Batch (Ref. Table 2.4-2) (Ref Table 2.4-1) (Ref Tao 1= 2.4-7, O. Three Gallon Pail Pellet Hard or Green Scrap Geometry Mass /Modenstion Control (Ref. Table 2.4-2) (Ref. Table 2.4-8)

2. Cnidation Furnace High and Low Grade Scrap Geometry Volume Mass Control Roats and Sludge (Ref. Table 2.4-1) (Ref. Table 2.4-1)

Mass

I

3. Dissolvers High Grade Scrap Batches Geometry '

(Ref. Table 2.4-3) (Ref. Table 2.4-7)

4. Vessels Scrap Dissolving and Geometry Mass Control Density Control Solvent Extraction Process (Ref. Table 2.4-5) (Ref. Table 2.4-7) (Ref. Table 2.4-8)

Liq 11ds S. Leachers Low Grade Scrap Geometry Mass Control (Ref. Table 2.4-4) (Ref. Table 2.4-8)

Solvent Extraction Organic and Aqueous Geometry Density Control 6.

Liquid (Ref. Table 2.4-6) (Ref. Table 2.4-8)

7. traste Tank Aqueous and Nitrate Wastes Nnsity Control Density Control Density Control V-103

- - - - Sulvent Extraction -ICAP Process

~

Process Liquids and Scrap Geometry Mass, Density 8 .~ M iscellaneous Equipment (Ref. Table 2.4-11) as Applicable Page 2.4-34

Reference:

SNM-1097

June 1, 1984 l

l pails have identified secondary controls depending on

-the contents. These are as follows: i Three gallon pails of pellets or hard scrap are critically safe because of mass / moderation control:

Mass-control because 35 kg is less than 90% of a minimum critical mass, and moderation control because hard scrap (pressed and sintered) contains significantly less than 0.5% by weight of water. (As shown in Table 2.4-8, the Kinf for UO 2 powder and 0.5% by weight of water is less than 1.0. Heterogeneous effects are unimportant at this level of water.) .For green scrap (i.e., unsintered pellets) the moderation content can be higher (as much as 5% equivalent water by weight) but green scrap occupies at least twice the volume of hard scrap.

Five gallon pails with 00 2 powder, wet sludges or low grade scrap are critically safe by virtue of volume.

The volume of a five gallon pail is about 22 liters . ;

which is. less than the 24 liter safe geometry volume fod homogeneous UO, and water mixtures.. This is a differeqt form of control from the geometrical safety discussed above because in the event of most failures of the pails geometry (diameter or height), the contents will still not exceed a safe volume. In those cases in which the safe volume may be compromised, the five gallon pail will still be critically safe because of tertiary control (mass / moderation control for pails of_high grad 4 powder and safe. batch control for wet and low grade scrap). Mass / moderation control is based upon a combination of mass (35 kg U0 2 is less than 90% of a minimum critical' mass for 5.0% enriched U02 Powder) and moderation contr-1 (U0 2 powder with powder preparation additives contains less than 5.0% by weig.ht of moderator equivalent to water). Wet and low grade herap is critically safe by safe batch control becLuse wet sludge is nominally 50% by weight of water (504 R 35 kg = 17.5 kg)-and pails of low grade scrap such as incinerator ash

!' contain less than 17.5 kg of uranium. (As noted in Table kg).

2.4-7, the safe batch limit for'U0, 2 powder is 18.1 t

2.4.3.2.2 Oxidation furnace boats are uded for the oxidation of high and low grade scrap and sludgns originally contained in three or five gallon pails. Individual boats are filled one-for-one from individual pails so i

Reference:

SNM-1097 Page 2.4-35 L

June 1, 1984

, that the secondary (and tertiary) controls for three and

, five gallon pails apply in the same way to the furnace 4

boats. The primary control is different, however, in that the thick by inch wide by inch t

~1ong furnace boats are geometrically safe for both homogeneous and heterogeneous UO 2 and water mixtures.

(Geometric safety for heterogeneous mixtures is evident since the buckling for a inch by inch x t

' inch slab, assuming an extrapolation length of 6.64 cm, is 0.0252 cm2 and is greater than the value of 0.0196 cm 2 for a inch thick safe geometry infinite t slab.)

2.4.3.2.3 Each dissolver processes kg UO 2 equivalent ,

batches of oxidized high grade scrap (typically U 0 3 8 powder). Table 2.4-3 illustrates that these vessels are geometrically safe with a Reff less than 0.90 for 4

, heterogeneous UO2.and water mixtures. The key features of the dissolver geometry are the dimensions of the - L vessel.and neutron absorber panele. This is the basis for the secondary control for tne dissolvers. Failure of the primary control can occur in two ways: f ailu're of the. vessel, or failure of the neutron absorber

, panels. In the case of the former, the failure will result in leakage of the dissolver solu tion on the floor ,

, into a safe slab configuration. If the neutron absorber i panels fail, however, the dissolver will be critically safe because of the batch size. The kg batch size t i - --

---~is less than the dissolver safe batch as tabulated in e

4 Table 2.4-7. It can also be noted that the uranium concentration in the dissolver after complete dissolution of the kg batch is 350 grams U/ liter. t From a comparison of. Tables 2.4-3 and 4-4, it can be seen that the keff of an unfailed dissolver is less than 0.83 -(since 350 grams U/ liter is equivalent to a fuel mixture with a W/F volume ratio of' greater than 6.0).

2.4.3.2.4 Vessels (as opposed to dissolvers) in the scrap process operation are used as holding or treatment vessels for uranyl nitrate from the dissolving or solvent extraction operations. As indicated by Table 2.4-3 the primary criticality control is geometry (again based upon- the radial dimensions of the tank and neutron t absorber panels). The secondary control for these vessels is either mass (reference Table 2.4-3 or l

Reference:

SNM-1097 Page 2.4-36 J

,,- - , - - , , - . ~ < , - - ~ ~ - - - - - - - . . , , , - --,---,r -- --------,n---- -

u--. ,,--wm ,,--m-,n

June 1, 1984 concentration density (less than 250 grams U/ liter, reference Table 2.4-5).

An exception to this is the secondary control applicable to the dissolver surge tank and-SX feed t adjustment tanks. The uranium concentration in these vessels can reach a maximum of 350 grams uranium per liter (the maximum value for the dissolvers) and the vessels'can contain as much as two dissolver batches

~(i.e., kg). Secondary. control of criticality t safety is therefore based upon the form of the material

-(uranyl nitrate) and a consideration of credible accident conditions. The worst case of these accident conditions are failure of the tank (such that the uranyl nitrate is discharged into the dissolver area basin and sump) and failure of the neutron absorber panels. . Criticality safety under either of these conditions is demonstrated by the applicable parts of Chapter 4. ;

2.4.3.2.5 Residue leachers are inch schedule t stainless steel 304 cylindrical tanks which are used in dissolving single three or five gallon pails of low grade scrap. These tanks are geometrically safe for heterogeneous UO 2 and water mixtures (reference Table 2.4-4) which is the primary control, and are controlled secondarily by mass (one five gallon pail.of low grade scrap contains less than one safe batch).- A. tertiary control is density since the concentration after dissolution is less than the minimum critical value for UO 2 and water.

2.4.3.2.6 Solvent extraction columns are geometrically safe (reference Table 2.4-6) - the primary control - and are controlled by concentration - the secondary control. As noted in Section 2.4.2.6, the maximum concentration accommodated by the organic phase in the solvent extraction columns is 125-130 gram U/ liter which is less than the 250 gm U/ liter value for U0 2 and water which corresponds to Kinf = 1. In addition, the aqueous phase (feed) concentration at the feed adjustment tanks is controlled to . gm U/1. t 2'.4.3.2.7 The nitrate waste tank V-103 is used to hold dilute effluent (< ppm U) from primary nitrate waste and t spent solvent extraction aqueous waste containing less than ppm U. The primary criticality control for t l

l 1

Reference:

SNM-1097 Page 2.4-37

June 1, 1984 this tank is the concentration control based on the solvent extraction process and process control system.

The secondary control is also concentration, in this case provided by the UPMP lab measurement of the uranium in the solvent extraction aqueous waste at the aqueous waste recirculating and surge. tanks. The -lab measurements are provided to the process control system. A tertiary density control is also present in the nitrate waste tank in the form of density probes and tank recirculation. As in the case of the fluoride waste system surge tank V-106, the nitrate waste tank density probes are directly interfaced to the process control system and are set to close off inlet streams if the tank bulk density exceeds a value equivalent to 250 '

grams U/ liter. This value is higher than the corresponding control in the fluoride waste system since densities in V-103 are significantly affected by metals and other impurities resulting from the solvent extraction operations. .

i, 2.4.3.2.8- Other miscellaneous equipment in the scrap processing operation are listed in Table 2.4-11. In this table, the primary geometry control is' described and the secondary controls are either mass or concentration controls as applicable to the system to which the i component belongs. 1 2.4.3.3 Geometry Control of the Scrap Processing System -

The scrap processing system was designed using geometry control as a guideline for the entire process. The equipment used in scrap processing can be summarized by the following:

(1) geometrically safe vessels t l (2) geometrically safe-tanks t (3) geometrically safe pumps t (4) geometrically safe heat exchangers t (5) geometrically safe filters t I

l Page 2.4-38

Reference:

SNM-1097

r O

June 1, 1984 (6) geometrically safe feed hoppers t (7) geometrically safe feed valves t (8) geometrically safe cooler t (9) geometrically safe slab filter t (10) ,

existing critically safe tanks t (11) geometrically safe process columns t (12) critically safe venturi eductor t (13) Material handling equipment geometrically designed for criticality safety (14) geometrically unsafe tank t (15) Criticality safe floor basins for liquid spills '

(16) Associated geometrically safe process piping Page 2.4-39

Reference:

SNM-1097

. - . - . - _ _ . - - - _ - - ~

June 1, 1984 r

This scrap processing system equipment is delineated in Table 2.4-11.

2.4.3.4 Normal and Accident Conditions in the Scrap Processing Operations As noted in the previous section, uranium processing equipment in the UPMP scrap processing operation have been designed to assure criticality safety under ununsual and. extremely unlikely conditions and are usually significantly less reactive (from a criticality safety viewpoint) than the optimum conditions. An example of this is that while most scrap processing operation equipment is designed ~ to be geometrically safe 8 (which implies optimum moderation UO 2 equivalent densities of 2.0 to 3.0 grams /cc), the maximum credible concentration in most operations is 350 grams ;

U/ liter = 0.4 grams UO 2 /cc. Table 2.4-12 presents a further criticality safety evaluation of the normal -

3,

operations for the eight scrap processing operation categories discussed in the previous sections. This table summarizes the normal conditions applicable to the category and indicates an " upper limit" Keff 2 3a value which can be assigned based upon the indicated reference. The upper limit so indicated is not y

. necessarily the factor which applies'to the normal i J

condition but is a value which by inference is greater than the actual value. These normal case Keffs are less 4

._.than 0.90.

In the evaluation of the scrap processing operations

! accident conditions, four sign'ificant conditions have ,

been' identified. These are:

2.4.3.4.1 Failure of Geometry Control in Scrap Process operations Criticality safety in the event of failure of geometry control in scrap process operations has been addressed in Section 2.4.3.2 and especially by Table 2.4-10. The secondary controls assure criticality safety when the i primary geometry control fails.

i 4

4 i Page 2.4-40

Reference:

SNM-1097

~

June 1, 1984 TABLE 2.4-11 SCRAP PROCESSING SYSTEM EQUIPMENT LIST Equipment Normally Label Contains criticality Safety Equipment Designation Uranium Geometry Description Description Yank Yes Geometrically Safe Cylinder t t

t Column Yes Geometrically Safe cylinder t t

t t

Column Yes Geometrically Safe Cylinder t t

t t

t Column Yes Geometrically Safe Cylinder t t

- t

', t t

Filter Yes Critically Safe Slab t

. t t

t Dissolvers Yes Geometrically Safe w/ Neutron t Absorber Panels t t

t:

Leachers Yes Critically Safe Cylinder t' t'

Leacher Yes Geometrically Safey Cylinder t t

Furnace Scrubber Yes Geometrically Safe Cylinder t it t'-

Condenser No Geometrically Safe Cylinders t; t.

t:

t' Cooler Yes Geometrically Safe Cylinder t!

t*

Cooler Yes Geometrically Safe cylinder t.

Condenser No Geometrically Safe. Cylinder t

t t

Cooler Yes Geometrically Safe Cylinder t t

Nester No Geometrically Safe Cylinder t t

t Yank Yes Geometrically Safe Cylinder t t

Page 2.4-41 Reference! SNM-1097

June 1, 1984 TABLE 2.4-11 SCRAP PROCESSING SYSTEM EQUIPMENT LIST '

Page 2 Normally Equipeent .

Contains criticality Safety Label Equipment Designation Uranium Geometry Description Description Concentrator No Geometrically Safe Cylinder t t

t Concentrator. Yes Geometrically Safe Slab and t Volume t t

Concentrator No Geometrically Safe Cylinder e

Concentrator Yes Geometrically Saft cylinder t t

t Filter No Geometrically Safe Cylinder t t

Filter Yes Seometrically Safe cylinder t

t Dust Collectors Yes Geometrically Safe Cylind'er t t

t Filter Yes Geometrically Safe Cylinder t t

t l

Filter Yes Geometrically Safe Slab t I

t

+

l Filter Yes Geometrically Safe Cylinder L

3-t Filter No Geometrically Safe Cylinder t i

t s

Yes Strainer Geometrically Safe Cylinder Filter Yes Geometrically Safe Cylinder t

-t Filter No Geometrically Safe Cylinder t!

tl Filter. No Geometrically Safe Cylinder t' t

I Filter No I

Geometrically Safe Cylinder t t

t Filter No Geometrically Safe Cylinder i t

Furnace Yes , Administrative Control t t

t t

8 Boats Yes Geometrically Safe Slab and t volume t t

Skids Yes Physical Design Maintains ?

Center-to-Center 25' Pall t Spacing t t

Transfer Dolly Yes Physical Design Maintains 25' t Center-to-Center Pall Spacing t t

Scrubber Eductor Yes Geometrically Safe Cylinder t t

Pump Yes Geometrically Safe Cylinder t t

t Page 2.4-42

Reference:

SNM-1097

June 1,.1984 TABLE 2.4-11 SCRAP PROCESSING SYSTEM EQUIPMENT LIST Page 3 Normally Criticality Safety Equipment contains Description Feuipment Designation Uranium Ceometry Description Label _

Yes Geometrically Safe Slab and t Pump ,

Volume t t

No Geometrically Safe Slab and t Pump volume t t

No Geometrically Safe Volume i Pump t t

Yes Geometrically Safe Slab and i Pumps Volume t t

Yes Geometrically Safe Volume t Pumps t t

No Geometrically Saf e Slab*and i

Pump Volume t t

No Geometrically Safe Slab and t Pump volume t t

Yes Coometrically Safe Volume t Pumps t t

Geometrically Safe Volume t Pump .Yes t t

Yes Geometrically Safe Slab and t Pump volume t t,

Yes Geometrically Safe Slab and ?

--- -- Pomp Volume t, t

No Geometrically Safe Volume i

- Pump t t

Yes Geometrically Safe Volume t Pump t t

Yes Geometrically Safe Slab and t Pump volume t t

Yes Geometrically Safe cylinder t Pump t t

Yes Coometrically Safe Slab and t Pump Volume t t

Yes Coometrically Safe Volume t Pump t t

No Geometrically Safe Slab and t Pump Volume t t

No Geometrically Safe Slab and t Pump volume t t

Yes Ceometrically Safe Slab and i Pump volume t t

Page 2.4-43

Reference:

SNM-1097

June 1, 1984 TABLE 2.4-11 SCRAP PROCESSING SYSTEM EQUIPMENT LIST Page 4 Normally Equipment Contains Criticality Safety Equipment Destenation Uranium Geometry Description Description Label Pump , Yes Geometrically Safe Slab and t volume t t

Pump No Geometrically Safe Slab and t volume t t

Pump No Geometrically Safe Volume i

f Pump Yes Geometrically Safe Slab and t Volume t t

Pump No Geometrically Safe Volume t t

t Pump Yes Geometrically Safe Slab and t volume e t t

Pump Yes Geometrically Safe Slab and t Volume t t

Pump Yes Geostrically Safe Slab and t volume t t

Pump Yes Geometrically Safe Volume i

Pump Yes Geometrically Safe Vcluse t t

. Pump Yes Geonetrically Safe Volume t.

t No Geometrically Safe volume Pump ,

t Yes Geometrically Safe Volume t Pulsers t t

Yes Geometrically Safety Volume t Pulser t t

notary valve Yes Geometrically Safe Cylinder t t

t Yes Geometrically Safe cylinder t notary valves and volume t t

t No Geometrically Safe Cylinder i Seal Pots t t

No Geometrically Safe Cylinder i Seal Pot t No Geometrically Safe Cylinder t Sump t No Geometrically Safe cylinder t susp t No Geometrically Safe Cylinder i Sump e No Geometrically Safe Cylinder i Sump t Page 2.4-44

Reference:

SNM-1097

June 1, 1984 TABLE 2.4-11 SCRAP PROCESSING SYSTEM EQUIPMENT LIST Page 5 Normally Equipment Contains Criticality Safety Label teuipment Designation Uranium Geometry Description Description Sump No Geometrically Safe Cylinder t t

Tank Yes Geometrically Safe Cylinder t t

t Tank No Geometrically Unsafe t

t Mopper Yes Geometrically Safe slab t t

Hoppers Yes Geometrically Safe Cylinder t t

t Tank No Geometrically Safe Cylinder - t t

t

, Tank No Geometrically Safe Cylindge t

, t t

Tank Yes Geometrically Safe Cylinder t t

t Tank Yes Geometrically Safe Cylinder i t

Deelsters Yes Geometrically Safe Cylinder t t

Demisters Yes Geometrically Safe Cylinder t t

t Tanks Yes Critically Safe Vessel t t

-Dentster Yes Geometrically Safe Cylinder t t

Tank Yes Geometrically Safe Cylinder t t

t Tank No Geometrically Safe cylinder t t

t Tank No Geometrically Safe Cylinder i t

t Tank Yes Geometrically Safe Cylinder t t

t Tank No Geometrically Safe Cylinder t t

Geometrically Safe Slab t Tanks (tsisting) Yes t t

t t

t Tank No Geometrically Safe Cylinder t t

Tank No Geometrically Safe Cylinder t t

Tank Yes Geometrically Safe Cylinder t t

Decanter Yes Geometrically Safe Cylinder t t

t Tanks Yes -Geometrically Safe cylinder t t

Page 2.4-45 Reference SNM-1097

June 1, 1984 TABLE 2.4-11 SCRAP' PROCESSING SYSTEM EQUIPMENT LIST Page 6 Normally Equipment Contains Label Equipment Designation Criticality Safety Uranium Ceemetry Description Description Decanter Yes Geometrically Safe Cylinder t t

Tank Yes Geometrically Safe W/ Neutron t Absorber Panels t s

Tank Yes Ceometricall Safe W/ Neutron Absorber Panels .'

t t

Tank Yes Geometrically Safe w/ Neutron ?

Absorber Panels t t

t Tank Yes Geometrically Safe W/ Neutron t Absorber Panels . i t

t t

Tanks Yes Geometrically Safe W/ Neutron t Absorber Panels t t

t Tank Yes Geometrically Safe W/ Neutron t Absorber Panels t 1

Tank Yes Geometrically Safe w/ Neutron l Absorber Panels t t

t

_ . Tanks Yes Geometrically Safe W/ Neutron t Absorber Panels

. Tanks Yes Geometrically Safe w/ Neutron t Absorber Panels t t

Pive (5) Callon Pails Yes 11-1/4* 10, 13-1/2* High 22 Geometrically Safe Volume at Furnace Discharge Liters Floor Basins No Mas. 3-3/4" Thick Slab Geometrically Safe Slab With Concrete and 1* Water Reflection Process Lines Yes 1/2* tc 4* Range Geometrically Safe Cylinders and Slabs Three (3) Callon Pails Yes 11-1/4* 10, 8-13/16* High 17.8 Geometrically Safe Volume at Purnace Discharge Liters Page 2.4-46

Reference:

SNM-1097

June 1, 1984 TABLE 2.4-12 SCRAP PROCESS OPERATIONS NORMAL CONDITIONS Upper Limit Equipment Category Normal Conditions gett 2 o Reference

1. Fails

- Five Callon High or Low Grade Scrap 0.8672 0.0041 Table 2.4 2

(< 10% H2O)

Wet Sludge Q 50% H 2O) 0.8485 0.0044 Table 2.4-2

- Three Gallon Dry Hard Scrap or Pellete 0.8362 0.0033 Table 2.4-2

(> W/F Ratio of 1.0)

2. Oxidation Furnace UO Powder or Sludge 0.84* -

Table 2.4-1 SakeGeometry-3-3/4" Slab ,

Dry Hard Scrap or Pellete 0.84 --

Table 4-38 '

i

'(< W/F Ratio of 1.0)

3. Dissolvers < 350 grams U/ Liter Table 2.4-3 t Tor W/r Ratio n 6.0) or 250 Kg U Mass Limit

4. vessels 1 350 grams U/ Liter Table 2.4-5 t

5. Leachers Safe Batch or < 100 Table 2.4-4 t grams U/ liter

6. Solvent Extraction < 350 grams U/ Liter Table 2.4-6 t T 50s n2 0)

7. Nitrate Waste Tank i 100 ppm U Table 2.4-8 t

8. Miscellaneous Types safe Geometry --

Table 2.4-1 Critically Safe Density Control See Chapter 4.0 Equipment < Mass Control See Chapter 4.0

From the Reactivity Formula and ARM-600 Mint, M2 and 1 Data, s

Page 2.4-47

Reference:

SNM-1097

June 1, 1984 2.4.3.4.2 Double Batching in Scrap Process Operations In scrap process operations, batch (or mass) control is only applicable to geometrically safe equipment (three and five gallon pails, oxidation furnace boats, leachers and various miscellaneous items). Therefore, double batching or failure of mass controls will not by itself result in a critically unsafe condition.

2.4.3.4.3 Failure of Concentration Control in Nitrate Waste Tank V-103 Since failure of concentration control in tank V-103 requires at least three unrelated, simultaneous -

incidents criticality safety of this vessel complies-with the double contingency policy. As discussed in i

Section 2.4.3.2.7, all of the following are required for failure of concentration control in V-103:

e Failure of solvent extraction process. controls '

(which normally limit the uranium content in the aqueous waste stream to ppm). t

Failure of the UPMP lab measurement system and UPMP process control system (which restricts dumps to V- -

103'from the nitrate waste system to no more than ppm U and from solvent extraction to no more t than ppm U). t close all inlet valves if tank bulk densities equivalent to greater.than grams U/ liter are t detected).

2.4.3.4.4 Uranium Backflow to Geometrically Unsafe Process vessels The process chemical demand in the facility is at a level which cannot be accommodated by the volume contained in 10" diameter-and smaller tanks without requiring unreasonable tank heights. Geometrically unsafe standard chemical make-up tanks are used to provide the required process chemicals.

Page 2.4-48 Referenc?:. SNM-1097

June 1, 1984 2.4.3.4.4.1 General Backflow Prevention Techniques In the facility design described in this document, three basic concepts (break tanks, air breaks and double block and bleed valving systems) have been utilized either individually or in combination to prevent backflow of process liquids to the geometrically unsafe tanks.

2.4.3.4.4.1.1 Break Tanks As used in this design, break tanks are geometrically safe 10" diameter, Schedule 80 pipe tanks or smaller.

The inlet supply nozzles from the geometrically unsafe chemical tank are locatdd in the top of the break tank, well above the break tank overflow. This assures that there is always an air gap between the liquid level and the inlet supply nozzle. The feed to the process vessel is from the bottom of the tank and the liquid level is controlled by a level switch interlocked with the . ;

.. chemical supply pump.- '

2.4.3.4.4.1.2 Air Break The air break is standard throughout this facility. It is used either alone or in conjunction with the other back flow prevention schemes. In the air break design, the inlet nozzle of the process vessel is positioned in the top of the vessel six to twelve inches above the vessel overflow nozzle. This positioning technique assures that an air gap always exists between the liquid in the process vessel and the supply line from the chemical supply tank (which at times is a break tank).

2.4.3.4.4.1.3 Double Block and Bleed Valving System The double block and bleed system was previously described in Section 2.1.3.4.3 and in Figure 2.1-8.

This concept is used where positive system pressures make the provision for .in air gap either in the break tank or the process vesrel unattainable. The general concept, as described in Figure 2.1-8, is to always have the supply tank isolated from the process by valve seats. This is accomplished by providing a drain path between the valves for unintentional flow from either the process chemical tank or the process vessel itself.

Reference:

SNM-1097 Page 2.4-49

June 1, 1984 2.4.3.4.4.2 Scrap Processing Operation Backflow Prevention ;

In the scrap processing operation, break tanks and air breaks are used to prevent process backflow. The ',

specific process chemical streams are:

(1) Nitric Acid All nitric acid supply to the dissolvers and leachers are via the nitric acid supply pump through the 8" diameter, Schedule 40, nitric i acid break tank to the process vessel. The supply *

' input in each dissolver/ leacher is accomplished '

across an air break arrangement.

, (2) De-Ionized Water De-ionized water is supplied to the system by the ;

DI water supply pump through the 10" diameter, '

t

. Schedule 80, DI water break tank into the process

[ vessels using the air break feature on each 4

t dissolver/ leacher.

(3) Aluminum Nitrate ;

1 Aluminum nitrate is fed to the system by the aluminum nitrate supply pump through the 10" i diameter, Schedule 80, aluminum nitrate break tank, into the process vessels which incorporate the air break design feature.

, (4) Solvent i

The fresh solvent is supplied from the solvent ,

} make-up tanks, which are not geometrically safe.

When a complete solvent replacement is required by '

) the process, the solvent is pumped via the solvent transfer pump through an' inter-tie pipe which i

requires administrative approval prior to

, connection to the solvent feed tank. This input is r i

made through the top of the vessel using an air break for back flow prevention. Additions are msJe '

in this fashion infrequently, and only when batch solvent concentration changes are required.

2. r l

y Reference SNM-1097 Page 2.4-50 l

i e -e,-,- . - o e-, ww n- w,-ww-e-- ,,-,.--m -w,-,--w mw-- - - , - . . ..- - - - , , , ---w-- ,---,e--,--n--n-- -, , - , - - - ,,,--,-,,,.,,,w, w,m, w,,, , ,,,-w., w--,n,,,a

June 1, 1984 During normal day to day operations, solvent make-up is from the geometrically unsafe solvent mix tank via the solvent transfer pump into the geometrically safe metering tank (break tank).

This tank is located six (6) feet above the solvent surge tank which it feeds by gravity flow. The input to the surge tank is through a nozzle located in the top of the tank, six inches above the tank overflow thereby forming the air break.

2.4.3.4.4.2.5 Aqua Ammonia Aqua ammonia is transferred to the 8" diameter SX neutralization tank from the main header through the top of the tank air break.

l Page 2.4-51

Reference:

SNM-1097

_ _ _ _ - _ - . = . - . . - . . - .

June 1, 1984 -

I 2.5 URANYL NITRATE CONVERSION 2.5.1 Current conversion Process j

The current Uranium Purification Process (UPS) is i

designed to process discrepant uranium powder with low impurity concentration. The system processes the powder through digest, precipitation, centrifugation and calcination steps to produce a uranium oxide powder. Refer to Figure 2.5-1.

2.5.1.1 Digest

' A weighed quantity of discrepant uranium powder (about kg) is dissolved in a controlled volu'me of t N nitric a'cid and then diluted with DI water to t l attain a uranium concentration of about grams / t liter. The digester is a inch diameter cylindrical t tank. The solution is transferred, through filters to ;

the slab storage (LEM) tank which are gallon, '

t inch slab tanks. There is a bank of LEM t tanks which are used for in-process storage of UNH i

solution ahead of the precipitation step.

2.5.1.2 Cooling and Precipitation

! A controlled volume of UNH is batch transferred from' a i LEM tank to a inch diameter cylindrical cooling t tank. The solution is cooled by recirculating through a heat exchanger. Ammonium hydroxide is added to obtain a predetermined pH level. When the solution is cooled to

. the desired temperature, it is transferred to a i inch diameter cylindrical tank where hydrogen t peroxide is added to precipitate the uranium as a i

_ tetroxide. The uranium is precipitated at a controlled f

pH as a tetroxide to prevent co-precipitation of iron and other metal impurities. The slurry is transferred

to a inch diameter cylindrical tank which is the t i

feed for the centrifuge.

l 2.5.1.3 Centrifuge The dilute slurry flows from the feed tank through the scroll type horizontal centrifuge. Here the uranium I f

1 1

Reference SNM-1097

June 1, 1984 FIGURE 2.5-1 UPS PROCESS - SCHEMATIC FLOW DIAGRAM t

t t

- t t

t

. t

. t t

t t

t t,

i t

. _ t t

t t

Page 2.5-2 References SNM-1097

June 1, 1984 precipitate is separated from the liquid phase. This is discharged into a slab type hopper as a thick paste.

The liquid phase is discharged into a second clarifier type centrifuge to remove the last' trace of uranium precipitate. The liquid phase centrate is discharged to the nitrate quarantine tanks. The solid phase, which is a slurry, is pumped back into the process through the centrifuge feed tank.

2.5.1.4 Calcination The slurry that is collected in the centrifuge hopper is pumped at a controlled rate into the calciner. The calciner is a rotating, gas heated inch diameter t tube through which the uranium slurry is processed. The slurry is dried and reacted at elevated temperatures with a counter current flow of hydrogen-nitrogen gas ;

mixture. The product is U0 2 Powder. The reacting '

gasses are withdrawn from the calciner, scrubbed with water to remove ammonia and nitrates and discharged into the plant process venting system.

The water from the gas scrubber is~ recirculated through a supply tank, cooled and reused in the scrubber. A controlled side stream is diverted to the centrifuge feed tank and blended with the process stream.

2.5.1.5 Quarantine The process water (nitrate waste, from the clarifier centrifuge) is alternately collected in gallon, t inch quarantine slab tanks. Each tank is filled, circulated for mixing and sampled for uranium concentration. If the uranium concentration is ppm or less, it is released to the t gallon storage tank, V-103, located outside the FMO building. If the uranium concentration is too high in the quarantine tank, che contents are recycled.

Another quarantine tank is used to collect acid flushing and rinsing of the ADU conversion lines. This was*.e solution is very acid since molar nitric acid t References SNM-1097 Page 2.5-3

June 1, 1984 is used to flush and clean the various process pipe lines and equipment. Ammonium hydroxide is added to the tank and recirculated for mixing to precipitate the uranium. The contents of this tank are circulated through a double centrifuge arrangement to remove the uranium solids until the concentration in the tank is ppm uranium or less, as determined by a SPEC 20 t Colorimeter. The tank contents are released and pumped to the nitrate waste storage tank V-103.

2.5.2 UPMP CONVERSION PROCESS With the implementation of solvent extraction, the uranyl nitrate solution will be a pure product. e

-Therefore the existing UPS conversion process can be modified to provide a t system. The conversion process can be t divided into the following steps: precipitation, centrifugation and calcination. . Figure 2.5-2 outlines ;

the process. The conversion process will be controlled '

by a central process control system similar to the one used for the other UPMP processes. '

2.5.2.1 Precipitation ' %l The UNH solution which,will be the product from the l solvent extr'actioir process, contains grams .t 6eanium per liter and about molar of free nitric t acid. This product, which will be the feed solution for -

UNH conversion'will be stored in existing gallon, i~nch'alab tanks in-the UPMP area. Upon t demand, the contsn'ts of one of the slab tanks will be

, transferred to the existing gallon LEM tanks, t 4

located in the FMO area. The precipitation will take

- c. place as shown in Figure 2.5-2. -

<~m .

4

'2 5.2.2 centrifuge ,

4

,w The solid bowl, scro'.1 type dewatering centr'ifuge will be an existirig ' uni 6 that was used for the UPS system. This unit separates the ammonium diuranate solids from the liquid. The solids will be discharged into a t

". < / ," .

i ~

] -

n . .

,.f., .

e *-

.i s ~ .

*9 *- *'

L

Reference:

SNM-1097 ',. ,

~.. -

June 1, 1984 FIGURE 2.5-2 UNH CONVERSION PROCESS - UPMP t

t t

t

. t t

t t

. . ; t t

t t

t t-t t

t t

t t-t t

t

. t t

t Page 2.5-5

Reference:

SNM-1097

t.

~

.?

3 i' ,

J n, ,

i' - -

June 1,.1984 E'

U , )

inch thick slab-hopper as a pasty sludge o.intainin'g about 4 solids and will be. pumped by the slurry t pump into the calciner.

2.5.2.3 Calcination Calcination will be accomplished in the same manner as current facility operations, The thick pasty sludge from the dewatering centrifuge will be pumped into the

.existisgi calciner. This sludge contains about 0: t

, solids and will be equivalent to about 4 uranium t concentration. The calciner will be an existing, '

rotating, gas fired 10 inch diameter tube, through which the uranium sludge will be processed.

2.5.2.4 Quarantine ,' s The process water from the clarifier; overflow tank will '

be pumped to the existing 230 gallon _ quarantine tanks.as nitrate waste. When a tank. is full, the liquid will be '

mixed by recirculation quarantinet. discharge, 'samplsd and analyzed for uranium using a SPEC'20 Colorimeter.- If the' uranium concentration is less'than the discharge l imit of, ppm, the tank contents will be released t and pumped to the UPMP nitrate waste collecting vessel -

for-further treatment. (See Section 2.3).

1-1 The operation of the_C nitrate qEarantine tank, as

- - --- described in Section 2.5.1.5, will' not change except\

that the contents will be transferred toathe UPMP nitrate waste collecting vessel. 1 2.5.3 Criticality safety considerations I s

,. s

~

The uranylinitrate conversion system, because of

~

-relatively high uranium concentrations, uses geometry

. control, operational process control, and chemistry i' limitations. A brief summary of cr'iticality safety limitc'and parameters which are applicable to the controls implemented in the UNH conversion process, are presented in Section 2.5.3.1. ~This includes a

~

discussion of the application of these ' limits and l '

parameters to the criticality safety considerations for /

\

\ t J

6

1

Reference:

SNM-1097 , Page 2.5 26

< ~, ;

1 s b s i 5, [

__-- m .-_._a. _ . . _

a f

June 1, 1984

, the process as based on the in-depth discussions hp' presented in Chapter 4. A final section is included fi which covers normal operations and accident conditions. ~

A major portion of the uranyl nitrate conversion Y. equipment was originally used for- the UPS

. conversion process. This includes the UNH storage

~'

tanks, the LEM and UPS quarantine tanks, the dewatering 4

, and clarification centrifuges, and the calciner. This equipment is currently approved for processing 4%

enriched uranium in the same density ranges required for

' UNH conversion and will be appropriately reanalyzed and

, modified, if necessary, prior to operations with enrichments above this value.

2.5.3.1 Concentration Control of the Uranyl Nitrate for the

, Conversion Process y

y , The uranium concentration of the feed to UNH conversion; will be controlled by the solvent extraction product '

concentration as defined in Section 2.4.2.7. The i a '

uranium concentration in each UNH storage tank will be known and verified by separat sampling and analysis by the Chemet laboratory and released by documentation from the Quality control and Production Control functions. Table 2.5-1 summarizes these concentrations at key points.in the conversion process and compares them with corresponding values for the UPS System. The chemistry-and process controls, which assure these

,s concentrations, are described in the following section.

2.5.3.2 Uranyl Nitrate Conversion Chemistry and Process Control 2.5.3.2.1 Chemistry The uranium concentration and the free nitric acid concentration for the UNH in each storage tank will be known'. These values are introduced into the process control system by the operator. Precipitation will.be controlled. Conversion will take place in a reducing atmosphere to product 00 2*

L , 2.5.3.2.2 Process Control' I

The UNH conversion process is controlled by a process control system. The control system will regulate the

.r %

?. ( ',

Reference:

SNM-1097 Page 2.5-7

, . - - , . ,-n.~,_. ---...---,_ _--. ,_,,,,.. ,.,,.__,. , ,-, , , - - . - .,.._-,.n n.---, -- . - - - - - - , . , . , - . , , . , , , - . . - , - - . - . , - , , .,

June 1, 1984 TABLE 2.5-1 UNH CONVERSION CONCENTRATION RANGE U Concentration Range t Prior to After Overall t System UPMP UPMP Effect t t

Digest Tank ---

t t

t UNH Recovery t t

LEM Tanks Inter- '

mediate .

Process t t

Intermediate ---

N/A t Process. t

. t Intermediate Inter- .

t Process mediate t Process t t

Intermediate ---

Inter- t Process mediate. t.

Process j Intermediate ---

Inter- t

_. Process __ ._.

mediate t Process t

Centrifuge Feed Improved t t

Sludge Unchanged t t

t Centrate Improved t t

t

'Centrate Improved t.

t

~

t Slurry Unchanged t t

t Scrubber Water Unchanged t t

t Quarantine Tanks Improved t t

Reference:

SNM-1097 Page 2.5-8

June 1, 1984 process flows, temperatures and pressures to produce the desired quality of UO 2 Powder.

2.5.3.3 Geometry Control of the Uranyl Nitrate Conversion System UNH Conversion w'ill be implemented with extensive use of geometry control in the design of the equipment. This equipment is summarized as follows:

(1)

Critically Safe Slab Tanks (existing) t J

-(2) Geometrically Safe Slab Tanks (existing) t

~(3) Geometrically Safe Heat Exchangers t I (4) Geometrically Safe Tanks t (5) Existing Critically Safe Centrifuges t (6) Critically Safe Slab Tank ! t

?

(7) .Etisting Critically Safe Cylindrical t

, Calciner (8) I;ritically Safe Heat Exchanger t (9). Geometrically Safe Pumps t

! (10) One (1) Critically Safe Floor Basin for-Liquid Spills (11) Associated Geometrically Safe Process Piping Table 2.5-2 lists the equipment associated with the UNH conversion system.

2.5.3.4 Normal and Accident Conditions in the Uranyl Nitrate conversion System As noted previously, the UNH conversion system will be a

modification of an existing process. This existing facility is critically safe for processing 4% uranium.

The modification will be the replacement of' existing batch type process equipment with new precipitation 1

4

}-

i i

Reference:

.SNM-1097 Page 2.5-9

June 1, 1984 TABLE 2-5.2 UPMP UNH CONVERSION EQUIPMENT LIST i

Normally Equipment Contains Criticality Safety Label Equipment Designation Uranium Geometry Description Description Centrifuge Yes Critically Safe Geometry

' t t

t Centrifuge Yes Critically Safe Volume t t

t Heat Exchanger Yes Geometrically Safe Cylinder and Slab ,

t Heat tachanger No Geometrically Safe Cylinder t and Slab t t

Heat tachanger Yes critically Safe Slab and t Volume t t

Calciner (esisting) Yes Geometrically Safe Cy1)nder t

. ?

Pump Yes Geometrically Safe Slab and t volume t t

Pump Yes Geometrically Safe Slab and ?

Volume t t

Pump Yes Geometrically Safe Slab and t volume t 1

Pump Yes Geometrically Safe Cylinder '1 i

Pump (existing) Yes Geometrically Safe cylinJer t t

'- Pump (esisting) Yes Geometrically Safe Slab and ?

Volume

Pump (existing) Yes Geometrically Safe Slab and ?

Volume t t

Pump Yes Geometrically Safe Slab and t volume t t

Pump Yes Geometrically Safe Slab and t Volume t t

Pump No Geometrically Safe Slab and t Volume t t

Pump (esisting) Yes Geometrically Safe Slab and t volume t t

Page 2.5-10

Reference:

SNM-1097

June 1, 1984 TABLE 2-5.2 UPMP UNH CONVERSION EQUIPMENT LIST (CONTINUED)

Normally Equipment Contains -

Criticality safety Label Equipment Designation Uranium Geometry Description Description Reactor Yes Geometrically safe Cylinder t t

?

Reactor Yes Geometrically Safe cylinder t t

t t

Tanks (existing) Yes. Critically Safe Slab t t

t Tanks (esisting) Yes Geometrically Safe Slab t t

Tank Yes Geometrically Safe Cylinder t

. ; t e t Tank Yes Geometrically Safe Cylinder t t

t Hopper (existing) Yes Geometrically Safe Slab t t

Tank (existing) Yes _ Geometrically Safe cylinder t t

t Tank (esisting) Yes Geometrically Safe cylinder t t

t Tanks (esisting) Yes Geometrically Safe Slab t t

t

~~' Scrubber (existing) Yes Geometrically Safe Cylinder t t

t Break Tank No Geometrically Safe Cylinder t t

t Tank (esisting) Yes Geometrically Safe Cylinder t t

t Break Tank No Geometrically Safe Cylinder t t

Process Piping ies 1/2" to 2* Pipe Geometrically Safe Cylinders and Slabs Floor Basin No 3* Thick Stab Critically safe Slab Five Gallon Pails Yes 11-1/4' 10 s 13-1/2' Migh, 22' Geometrically Safe Cylinders Liters

~

Reference:

SNM-1097 '*9*

June 1, 1984 equipment. This new equipment, as well as some of the existing equipment, will be geometrically safe for 5%

enriched uranium.

o ~ 2.5.3.4.1 Failure of Concentration Control As noted in Table 2.5-2, almost all of the equipment is geometrically safe and any failure of the UNH conversion concentrations control will not result in a critically unsafe condition.

2.5.3.4.2 Failure of Geometry Control j .

Failure of. geometry controlaof the UNH conversion system ,

will not result in a critically unsafe condition in new Uranyl Nitrate Conversion equipment because of the

! concentration control features described in Sections 2.5.3.1 and 2.5.3.2. Failure of geometry control in existing equipment is considered in the original ,'

, analyses for this equipment and will not be discussed' .

further in this document.

2.5.3.4.3 Backflow of Uranium into Unsafe Tanks All the process vessels used in the UNH conversion i system are geometrically safe. The flow of high I concentration material either upstream or downstream of I the process system will not result in an unsafe

___:.... condition since all those tanks and vessels are geometrically safe.

The. supply of and DI Water will be from geometrically safe break tanks which have air breaks from the main supply tanks.

4 T

e

{

Page 2.5-12

Reference:

SNM-1097

--, - - ,= _ _ . _ _ _ ..y,,.,,.,_.,,_,_,,,mm.. , , _ _ , , . , , , , , . , . . . . _ ~ , , . . _.,__m , , , , .

t June 1, 1984 1

l 2.6 PROCESS CONTROL- SYSTEM i

The UPMP process control system will be a commercially l

, available . system. It will t i

be a distributed control system with both batch and !

, continuous capabilities and will employ a t Figure 2.6-1 is a sketch of the t system configuration.

2.6.1 System' Power The process control system and laboratory equipment will be supplied with 120/208 VAC power through an Uninterruptible Power Supply (UPS) with a full load battery backup of at least 15 minutes. This UPS will be of the rectifier, battery charger, inverter and static switch type. The UPS provides power to the control system including field instruments, operators consoles, alarms, printers and loggers. It will not provide power 4 -

to operate field equipment. Field equipment and final '

control elements will be designed to "f ail' safe" upon loss of power. The UPS will be supplied power from the plant emergency power generator during plant power outages.

- 2.6.2. Control System Features The control system employs redundancy in three areas which are considered to require high reliability.

. 2.6.2.1 A will-be employed t even though the system will not be physically distributed. All control system controllers and input / output signals (I/0) will be located in one room 4

with operator consoles in the adjacent room. The will provide assured system t communications at all times as well as on-line maintenance of the communications system.

f 2.6.2.2 Dual operator consoles will allow dual control of each UPMP process and also provide for backup operation of all processes requiring high reliability by having those processes configured in both consoles. Key. switch operation will be used to prevent _ simultaneous operation of those processes at each console.

I Page 2.6-1 l

Reference:

SNM-1097

June 1, 1984 FIGURE 2.6-1 UPMP CONTROL SYSTEM t

t t

t

. ; t t

t t

+,

I t'

t t

t i, t t

- t t

t Page 2.6-2 References. SNM-1097

1 June 1, 1984 2.6.2.3 can be used in a t !

^

redundant configuration to provide high reliability for those processes configured on both operators consoles. ,

t When used in a redundant configuration, the will t j provide both continuous and t and sequencing operations. t

! will provide primary control with the t

in a standby state being continually t i

updated with current operating data from the t .

The will be ready to t

. assume control if the fails. Failures t

! will be detected by both internal and external cross-checking features.

). 2.6.3 System operation Signals to and from field devices will be interfaced to the control system through two types of highway connected devices. . ;

2.6.3.1 Analog controllers which will be used in continuous 4

control applications or which require complex control algorithms will be individual loop controllers. The mode of control (manual, auto, computer) and tuning l parameters (setpoint, gain, rate, reset) can be adjusted i

from the operator's console. Alarm contacts will be provided to allow communication of status type information between controllers without .using the data highway. Each controller will be a microprocessor that can be operated independent of highway integrity.

2.6.3.2 The . uses dedicated t I/O and its own programs to operate and control process L

units. The has its and can t be viewed as a with t extensive analog control capabilities t 2.6.3.3 control information other than start-stop and control parameter type'information will not be transmitted over the highway. Control takes place in the or t The and t

, , will use the highway to pass status and t process variables to the operators console for indication, trending, logging and operator interaction.

i i

4

Reference:

SNM-1097 Page 2.6-3

June 1, 1984 2.6.4 Operator Consoles

~Two operator consoles, each with two color CRTs, will provide real-time interactive graphics to display all process. parameters and provide the operator with the means to operate all processes and equipment, respond to alarm conditions, input requested data, demand CRT copies and logs, etc., through the use of numerical keys,' dedicated keys and " soft keys" whose functions are keyed to the CRT display. Each console will have a 4-position key switch: Locked, which allows only monitoring of all the displays; Operate, which allows the operator to perform all normal operating functionar Tune, which allows the shift foreman or process engineer -

to change process or control loop tuning parameters if permission to do so was input during configuration; and configure, which allows the control system engineer to modify or generate new graphic displays, control sequences, control logic, etc., through the use of a , ;

configuration keyboard on a separate engineering

ConMole.

A backup configuration keyboard will be supplied at one operator's console but will be covered and unavailable to the operator. Each console will have a capability ,

for 100 graphic displays which include plant overview, i groups, process graphics and special log formats. Each console also has the capability to do up to four real-time trends. Real-time and historical trending display for up to will also be available on'the console CRT through a trend unit on the data highway.

2.6.5 Hard Copy Printout Each operator's console.will support one printer and one logger. The logger will be capable of alpha-numeric printing only. The printer gives both alpha-numeric printing and black and white copies of CRT graphic displays on operator demand. ' Batch end reports, hourly, shift and daily logs or summaries, operation actions and alarms can be scheduled for printout on either the printer or logger as desired.

Reference SNM-1097 Page 2.6-4

l

. I i

June 1, 1984 CHAPTER 3.0 SUPPORT SERVICES & EQUIPMENT All processes involved in the Uranium Process Management Project require outside support and services. In most cases, the support systems provide chemicals required for operation which are not uranium bearing. .In cases such as the solvent treatment system, uranium bearing materials are treated but the system requires only a short description for clarity. Since there is a commonality of purpose, the- support systems have been grouped into one chapter ,

and are detailed in the following sections. ,

3.1 SOLVENT TREATMENT The solvent treatment system is designed to permit continuous treatment of the solvent inventory as it is circulated thro' ugh the solvent extraction process. The

, , basis of the process is the emulsification of the . ;

organic stream with an ammonium carbonate wash solution

thereby chemically scrubbing out the residual uranium and the mono- and di-butyl phosphate degradation

. products. The incorporationaof the continuous solvent treatment process allows batch acidification of the waste ammonium carbonate wash stream and adjustment of the solvent concentration during scheduled weekly shutdowns of the solvent extraction system.

The main solvent stream (CW) leaving the column, will contain approximately gm U/1 and will provide the t principal load to the solvent treatment system with minor flows coming via the aqueous waste decanter and from the vent system decanters. The CW stream flows from the strip column, to the spent solvent surge tank.

The solvent will be pumped from the spent solvent surge tank at twice the CW stream flow to the solvent wash tank where it will be combined with an equal volume of ammonium carbonate wash solution. The resultant two phase emulsion will be pumped from the bottom of the solvent wash tank through the solvent heater to the in-line mixer which takes suction from the upper region of the wash tank.

Page 3-1 References SNM-1097 ,

l l

l 4 June 1, 1984 l l

The intimately mixed phases will be routed to a liquid / liquid centrifuge, where the organic and aqueous ;

-phases are separated. The solvent phase from the i- centrifuge will be split, with approximately one-half being routed to the solvent feed tank,-.and the remainder being recycled to the spent solvent surge tank. The aqueous stream from the centrifuge will be recycled to

the chemical wash tank. Since the amount of impurities requiring removal by the solvent wash system are small, the aqQeous wash solution can be used for a week of operation without depletion of the ammonium carbonate solution.

-Treatment of the solvent with the ammonium carbonate wash solution will remove the hydrolysis products and their uranium complexes. The solubility limit for uranium in the form of ammonium uranyl carbonate is in the range of gm U/1. The reaction products from t

, solvent treatment strongly favor the aqueous phase and are effectively removed from the solvent. The aqueous ;'

wash solution is acidified on a batch basis during weekly shutdown of the solvent extraction system. The acidified aqueous phase will be centrifuged and disposed ;

to the nitrate waste treatment system for recovery of '

I residual uranium as a calcium uranate sludge. ,

Organics separated by centrifuging of the acidified wash

solution will be collected in the neutralization tank,

. _ _ . . . . _ . along with interfacial waste that is periodically purged t

from the solvent extraction columns. These organic bearing wastes are sampled for uranium and acid, neutralized with ammonium hydroxide, filtered and loaded into a drum for disposal via the incinerator.

3.1.1 Geometry Control of the Solvent Treatment System As with the major process systems discussed previously, the solvent treatment system design makes extensive use of geometry control. The following summarizes the equipment in use in the facility:

(1) geometrically safe vessel t (2) geometrically safe pumps t I

Page 3-2

Reference:

SNM-1097 1

June 1, 1984 (3) geometrically safe tanks t (4) geometrically safe heat exchangers t (5) geometrically safe centrifuge t (6) geometrically safe venturi eductor t (7) geometrically safe filter t Table 3-1 summarizes the process equipment along with a brief description of the geometry and corresponding criticality safety designation. '

3.2 . FILTER CLEANING STATION In both the nitrate waste treatment system and the rad waste treatment system, the filters may become plugged l with solide thereby restricting the flow rate. When ,

~

this occurs, it will be necessary to clean the filter i

media. This cleaning can be accomplished 'in two ways.

The first and most timely will be cleaning in place and l the second will be filter disassembly and transport to a l separate filter cleaning. area adjacent to the filter i

banks.

For in-situ filter cleaning, the media will be isolated from the process by closing the inlet and discharge valves. The filter housing will be drained. A epecial program in the UPMP process control system will be initiated and the filter will be flushed with dilute nitric acid.

If the in-situ cleaning is not effective, further cleaning of the filter will be necesary to maintain processing rates. The filter will be isolated, drained, and then_ disconnected from the piping system. The filter media will be transported to the cleaning station enclosure for further intensive cleaning.

Page 3-3

Reference:

SNM-1097

June 1, 1984 TABLE 3-1 SOLVENT TREATMENT SYSTEM EQUIPMENT LIST Normally Contains Criticality Safety agulpeent Description Laget 3,eutseent Destenation Urantus Geometry Description Centrifuge Yes Geometrically safe cylinder t t

t Nester Yes . Geometrically safe cylinder t e

Cooler No Geometrically safe cylinder t t

t t

Filter Yes Geometrically safe cylinder t t

t Eductor Yes Geometrically safe ey11nder t

t pump No Geometrically safe volume t t

Pump Yes Geometrically safe slab and t volume t t

pump Yes Geometrically safe slab and I volume I t

Pump Yes Geometrically safe slab and t volume t Tank No Geometrically safe cylinder

t Tank Yes Geometrically safe cylinder t t

t Tank Yes Geometrically safe cylinder t t

t Tank Yes Coometrically safe cylinder t t

t Tank Yes Geometrically safe cylinder t t

t Tank No Geometrically safe with t neutron absorter panels t S Gallon Pella Yes 11-1/4' 10, 13-1/2' hig%, 22 Geometrically safe volume liter volume Process Piping Yes 1/3' to 3' range Geoo g ally safe cylinders Page 3-4 Reference SNM-1097

l June 1,1984 In the enclosure, manual cleaning of the media will.be done with either a brush assembly or water jet wand.

All liquid discharge in the enclosure will be routed to the nitrate waste sump.

3.2.1 Geometry Control of the Filter cleaning Station Geometry control has been implemented in this system.

Table 3-2 lists the equipment summarized belows (1) geometrically safe cylinders t I

(2) geometrically safe pump t (3) Geometrically safe process lines 3.3 NO, ABSORPTION Oxides of nitrogen will be generated during the . ;

dissolving / leacher operations. The rate and compositioh of NOx generation will be dependent upon the rate of feed processing, the uranium oxidation state,.the quantity of metal contaminants (Fe') and the nitric acid j

strength used in the dissolution operation. -

The NOx generated in each dissolver in the scrap l processing operation will'be vented through two parallel, four inch diameter demisters. A DI water i

spray nozzle is provided for back wash of each individual demister pad. (Figure 3-1) i The two gas streams will be combined downstream of the reflux condensers where the condensibles are returned to the dissolver. The non-condensible stream will be combined with the non-condensibles coming from the leacher operation. This non-condensible stream from the leachers will go through a system similiar to that of the dissolvers.

The gasses generated in the pipe leachers t will be passed through a demister and to a 6-5/8" diameter leacher reflux condenser prior to mixing with the non-condensible gasses from the dissolving References SNM-1097 Page 3-5

June 1, 1984 TABLE 3-2 FILTER CLEANING STATION EQUIPMENT LIST Normally e sgulpeent containe criticality safety Leben equiseent Destenation Uranium Geometry Description Description pump No Ceometrically safe slab and t volume t

. ; t Tank No Geometrically safe cylinder i t

t Tank No Geometrically safe cylinder t t

process Lines No 1/2* to 2' Pipes Geometrically safe cylinders 1

Page 3-6 Referencet SNM-1097

June 1, 1984 FIGURE 3-1 DEMISTER PAD t

i t l

t t

. ; t

' t t

t t

t t-t t

t f .. -

t t

t o t

.t t

t t

t-t t

~ t t

.t t

l Page 3-7 References 8NM-1097 lr -

June 1, 1984 operation. These combined non-condensible gas streams .

will be passed through two parallel t to the NO x absorption column as shown in t Figure 3-2.

l.

l The NOx absorption. column is a inch in diameter t by foot tall tower whose active length.contains -t trays with bubble caps on each tray. The normal t absorber liquid is dilute nitric acid circulated through an external heat exchanger E-280. - The tower design does have the capability of operating using once-through de-ionized water as the . absorber liquid if necessary for

, improved efficiency. The NO x absorber is equipped with in-line instrumentation for monitoring tower a operating efficiency and effluent. concentration.

l The recovered acid will be returned to the SX process and DI make-up to the tower will be supplied from a separate source not connected to -the main process break; i tank. This acid use/DI replenish operation will '

prevent the build-up of any residual uranium particulate that may enter via the NO x off-gas stream from the dissolving operation.

3.3.1 Criticality Considerations

}

Since the effectiveness of any NO x absorption system will be based upon contact and chemical kinetics, it is

. not practical to design the absorber as a geometrically safe tower. Instead, criticality safety will be assured

. by designing a dissolver and leacher off-gas system which prevents the carryover of uranium to the

, absorption column and which permits the use of a conventional NOx absorber tower design. This will be i done by placing the gaseous vent (s) for each dissolver

or leacher at the top of the tank or vessel void space

!' and extending the straight vertical run beyond the tank overflow. Any particulate in the gas flow from the vessels will be trapped in either of the two parallel i demisters prior to going to the geometrically safe i'

reflux condanser which will further assure disengagement of any uranium. bearing liquids or particulate.

4 i

Reference SNM-1097 Page 3-8

,+-.-,,,,,r,--n,,,,c,- m-,1mm----.,m, ,,--,c.-,,,,, ,_n, y,w.mm,,,.y.--v _ m m e n .m--w+,m ~-- g_e-- m -,-.---.g.p,vv,--e----,

June 1, 1984 ;

i FIGURE 3-2 NOx ABSORBER SYSTEM t

t t

t 4

t t

t

, t t

i . t t

, t t

t t

t-t t

t I

t t-t t

4 t i

t 4 t t

t t

Page 3-9

Reference:

SNM-1097

. - - _ _ _ . . _ _ - . . . _ _ . _ _ . . . _ _ _ . _ . _ - . _ . . _ . . _ . _ . = _ . . _ . _ . _ - _ , . . _ . _ _ _ _ . , _

June 1, 1984 l

Since the gas flow from each leacher will be much lower than that for the dissolver, only one demister will be required. The final liquid disengagement will be done in a geometrically safe reflux condenser in the same manner as the dissolver effluents. In this sense, the

' geometrically. safe reflux condensers and demisters will' prevent the transport of solubilized or particulate i

uranium to the~NOx system and will be the basic criticality control for the NO x column. If this

redundant system were to fail, the liquid in the column sump section will be recirculated through a density element which will alert the operators to a major change in tower liquid composition.

p j 3.4 AQUEOUS AND SOLVENT MAKE-UP (ASMU)

The aqueous and solvent make-up system will be the primary chemical support facility for this project. The -

facilty will consist of six (6) independent chemical. :

j preparation processes combined with a storage and supply tank area to support the main process needs of the project. Due to the volumes of chemicals required, the support processes and supply vessels will be of conventional, geometrically unsafe design. Descriptions of the steps which will be taken to assure critically safety of the support area have been outlined under the I process descriptions of Chapter 2. In general, all '

criticality protection of these large support vessels will be accomplished by the use of either break tanks,

, air breaks, double block and bleed systems, or combinations.of the above.

Table 3-3 details both the processes included in the facility and the major storage vessels.

3.5 - SOLID WASTE HANDLING Combustible and non-combustible waste generated in the main process areas will be transported to a two ,

compartment waste box enclosure. The enclosure will accommodate two (2) (4' x 4' x 3 1/2') metal, wood, or cardboard boxes which will be placed in the enclosure through a rollup door. Each waste box will be monitored while in the enclosure to assure that the amount of

Reference:

TNM-1097 Page 3-10 l

I

i June 1, 1984 TABLE 3-3 AQUEOUS & SOLVENT MAKE-UP PROCESS

& CHEMICAL SUPPLY TANKS Processes 4

Aluminum Nitrate Make-up Weekly Solvent Adjustment Bulk Solvent Make-up/ Storage Ammonium Bicarbonate Make-up

, Ammonium Carbonate Make-up Storage and Supply Tanks

% Lime Storage Tank Gallons t

% Lime Storage Tank Gallons t DI Water Storage Tank 10,000 Gallons Rad Waste Tank Gallons t.

Reference SNM-1097 Page 3-11

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

1 June 1, 1984 uranium in the box is less than a safe batch. When the boxes are ready for removal, the box lids will be put on and the boxes transferred to the fuel manufacturing decontamination facility for further sorting prior to ultimate disposition to either the incinerator or burial.

3.6 PROCESS LABORATORY The UPMP process laboratory will occupy 680 square feet of space adjacent to the control room / office complex.

(Reference Figure 1-3). It will have thirty-four linear feet of open bench space, a sink and a fume hood with sink inside. Waste liquids will leave the lab through

drain piping either to rad waste (for the liquids containing only trace uranium and chemicals - e.g.,

glassware wash water, safety shower water, etc.) or to nitrate waste (if the uranium or chemical content is unknown or high - e.g., unused portions of samples, out*

of-date standards, etc.). Room air will be exhausted th the existing FMOX roof scrubber through ventilation provided for lab equipment and the fume hood. A normally closed rear exit door will be provided for emergency egress, and the room will be equipped with an overhead sprinkler system for fire control. Figure 3-3 i shows the schematic layout of the UPMP process I laboratory.

The laboratory will contain approximately fifteen pieces of newly purchased, standard analytical instruments and support equipment which supplement and serve as back-up to t These will provide selective, routine uranium t concentration monitoring from process locations by t performing analyses on samples t Both will be connected t to inch, which t to the various process sample points. t certain impurity measurements will also be made by the on additional inch t after the uranium has been removed by t systems. Each of t these inch t can be manually sampled t

Reference:

GNM-1097 Page 3-12

. ~ ..- - .. - .. -. - - __ - . . . .- . . - . - . . . _ - - . _

t i

i i June 1, 1984 1

l i

l FIGURE 3-3 UPMP PROCESS LABORATORY l l

! i 3

p i t

t i' 1 t t l t

i t .

t ,

t i

i t

t t

t

. t

, t ,

I ;

t

! t t l t

t i t t

! t !

I t

t I t j

t !

4 t ,

j t 1 t I I

t t i t

! t i t i t .

,1 t t !

t !

~

t o

6 s

i I References SNM-1097 Page 3-13 l i

f

June 1, 1984 in the lab for subsequent analysis on an or other instrument.

The process laboratory will be capable of performing all uranium and impurity analyses required for process control. Data from these analyses will be transmitted to the control room for use in process control activities. In addition, the lab will support the calibration checks of some of the on-line process measur'ement instrumentation.

i, 1

References SNM-1097 Page 3-14

m June 1, 1984

{

CHAPTER 4.0 i' CRITICALITY SAFETY ANALYSES ,

POR BQUIPMENT & OPERATIONS 4.1 GE-WMD CRITICALITY SAFETY POLICY & METHODS I The Ura'nium Process Management Project facility, equipment and operations have been designed in accordance with criticality safety requirements using analytical methods and techniques documented in WMD's l Special Nuclear Material License SNM-1097. The key !

elements are the following:

4.1.1 Double Contingency Policy Process designs shall incorporate sufficient factors of ,

safety to require at least two unlikely, independent and'.

concurrent changes in process conditions before a i criticality accident is possible. ;

4.1.2 Geometry Control ,

The preferred method for assuring nuclear criticality safety in production quantities of fissile materials is by the use of geometry control. Geometry control is defined to mean criticality safety under the following conditions:

4.1.2.1 optimum moderation by water or by any other material more effective than water which can credibly cause optimum moderation in the system.

4.1.2.2 The optimum credible density of the fissile material l taking into account the presence of the moderator as !

described in 4.1.2.1 above.

4.1.2.3 The presence of fissile material and moderator in all areas in which they are not physically excluded.

4.1.2.4 Full reflection by at least 12 inches of water at the i closest boundary to the equipment unless other more effective reflectors are actually present and in such a ,

configuration. (In this case, the more effective i reflector must be assumed). !

t l

1 Page 4-1 !

Reference:

8NM-1097 a_ _

, , /

i ?

s -

June 1, 1984

. , i

! - ~. f 4.1.2.5 The opTiinii d'egree3f credible Tieterogeneity, in fissile material, moderator ~and neutron absorbers.

4.1.2.~6

The maximum enrichment of Cissile material permitted in the system. ,.

~

4.1.2.? Maximud'and minimum equipment Af.mensienA, as appropClate, which are per,tinent to the geom,etrica1

+ configdration. ,

4.1.3 Fixed Neutron Absorbers .

a The use of. fixed neutron absorbe(i'is considered to be a '

form of geometry control to provide criticality safety under the following conditions .

4.1.3.1 'Theneutronabsorberisasolidhcableboroncompoun'J' ,

fixed in a matrix. -

i, 4.1.3.2 The neutron abs 6b'b'r e and any hydiogeneous material'used to thermalize neutrons are seat.ed'in a stainless' steel or other suitablefcontainst which'may be an integral

part of the proceas equipmen,t. ,

4.1. 3.* 3 : The neutron abs 0rher is installed as a permanent part of

~

I the equipment or system and is not easily moved or removed. ,

4.1.3.4 ' Prior to'its use, the presence of the neutron absorber is verified. ,

4.1.3.5 ,.The void volume of the neutron absorbet' system is negligibly sm(11 to prevent internal rearrangement.

4.1.3.6 The effectiveness of the neutron absorber system is analysed and demonstrated by use of validarod f calculational methods. ,

4.1.3.7 The integrity of the neutron absorber is mainthined aga' inst credible fire har.ards and corrosive conditions.

4.1.3.8 The integrity of the neutron absorber is v0rified on a periodic schedule compatible with the corrosive nature of the environment.

4.1.3.9 Equipment from which neutron absorbers are removed, is tagged out and not used.

.m f f References SNM-1097

I June 1, 1984 l 4.1.3.10 The neutron absorber is designed to withstand credible industrial accidents and natural events.

t 4.1.4 Administrative-controls Where geometric control is not practical, criticality l control may be based on control of U235 mass or control l of moderation.

4.1.4.1 Where control is based on U235 mass limits, process operations are limited such thatt i 4.1.4.1.1 The mass of any single accumulation does not exceed 45 percent of the minimum critical mass (defined to be a safe batch) or, 4.1.4.1.2 The mass of any single accumulation does not exceed l 75 percent of the minimum critical mass if double , ;

batching is not credible.

4.1.4.2 Where control is based on over moderation of fissile material accumulations, the accumulations are periodically assayed or sampled to assure the following:

4.1.4.2.1 The maximum fissile material concentration does not exceed one half of the minimum critical concentration for the given enrichment or 4.1.4.2.2 The H/U235 atomic ratio is not less than.5200 4.1.5 Definition of Critica11y' Safe The term " Critically Safe" means:

4.1.5.1 Under identified normal conditions, the effective l neutron multiplication factor (Keff) is such that:

l Reff + 3 o - Bias < 0.90 Where o is the statistical uncertainty (if any) in the l Keff value and bias is the applicable analytical bias based upon form'al validation of the method used to determine Reff. (Bias = calculated Keff - actual Reff) l Page 4-3 References SNM-1097 L

e - - .

June 1, 1984 ,

3 i 3-4.1.5.2 Under specified accident conditions (such as those specified in Section effective neutron 4.1.2 for geometry multiplication control)the factor satisfies l, Keff + 3 o - Bias < 0.97 4.1.6 Analytical Methods-Used In Criticality Safety Analyses The analytical methods used in criticality safety analyses are:

. 4.1.6.1 The'GEMER Monte Carlo Code'

.s ;

4.1.6.2 The GEKEy0 Monte Carlo Code with JRK Modified Hansen-Roach Cross Sections 4.1.6.3 The SAC Interaction Solid Angle Code 4.1.6.4 The reactivity formula with Kinf, M2 and A data taken from ARH-600 for homogeneous UO 2 and water mixtures.

4.1.6.5 The reactivity formula with Kinf, M2 and A data taken from DP-1014 and GEBLA calculations for heterogeneous lattices of 00 2 in water.

4.1.6.6 Standard geometric buckling formulas for cylinders, slabs and spheres.

\

4.1.6.7 Tabulated safe batch and safe geometry parameters Table 4-1 summarizes the analytical diases which are applicable to the Monte Carlo Codes >and reactivity formula. '

l 4.1.7 UO2 and Water Mixtures

( In criticality safety analyses, the optimum credible l form of fissile material and moderation is assumed to be I

full density UO2 and water mixtures. The term full density means that the UO2 and water combine such that i

the partial volumes they occupy add up to the total volume; that is, there is no void space or other " inert" material present. Full density homogeneous mixtures.can be conveniently characterized by the weight fraction of water in the mixture as: -s l

\ .

i

) ,

,' 3+

, Page 4-4

Reference:

SNM-1097 l

l 1

l l

h l June 1, 1984 l

+}

TABLE 4-1 BIASES FOR ANALYTICAL METHODS USED IN CRITICALITY 1

2 ., .

5 Method , Range of Application Bias Source-

1. GEMER Monte Lattices of Low -0.003 TRX Carlo Code Enriched U0 2 Rods in Benchmarks Water GEMER Monte All other Low Enriched -0.010 GEMER Carlo Code UO2 and Water Systems

Validation Study

? . ;

2. GEKENO Monte -Homogeneous Low 0.00 -Table 4-4 Carlo Code with Enriched UO2 and Water JRK Modified Systems Hansen-Roach Cross Sections GEKENO Monte Heterogeneous Low -0.01 Table 4-5 Carlo--Code with Enriched JRK Modified Hansen-Roach Cross Sections and adjusted U 238 Materials ids

3. Reactivity' Homogeneous Low -0.017 Table 4-4 Formula with Enriched UO2 and Water j ARH-600 Data Systems i ,.

[' 4. Reactivity Heterogeneous Low 0.00 Table 4-5 Formula with Enriched UO 2 and Water DP-1014 and Systems l GEBLA Data l

! ' .t ,.

s 3..

i' Page 4-5 l

Reference:

SNM-1097

.,g, _ _ , . _ _ ... , _ . . , -. _ - - - - - - - - - - - - - -

June 1, 1984 p-MIX = - - -

1- - - ---

WF UO, + WF-H;O-p(T)-UO2 p(T)-H 2O Where WF H 2 O is the weight fraction of water, WP UO2 =

1-WF H2 O is the weight fraction of UO 2, p(T)-UO 2 = 10.96

.gms/cm3 is the theoretical maximum density of UO 2' p(T)-Hf0 = 1.00 gms/cm 3 and p-MIX is the density of the UO2 and water mixture in grams /cm3 for these weight fractions. UO 2 -and water densities are easily obtained from this formula since:

p-UO2 - WF U02 x p-MIX and, p-H 2 O = WP U 20 x p-MIX , ;

(Table 4-4 shows typical results of the application of these relationships.)

For heterogeneous mixtures, characterization is usually ,

made by use of the water to fuel volume ratio (W/F). In i the case of square lattices of (infinite) unclad 00 2 rods in water, this value is :

W/F =~(L2- wD2)

.4 s D' 4

Where D is the diameter of the rod and L is the center to center separation between rods. This W/F Ratio is related to the WF H 2 O of a homogeneous Full Density UO 2 and Water mixture with the same total amounts of UO2 and water by:

WF H 2 O =

1 10.96 + 1 ,

W/F j (This is also illustrated in Table:4-4.) l Page 4-6

Reference:

SNM-1097

June 1, 1984 s 4.1.8 The Reactivity Formula The Reactivity Formula is often used to calculate effective' neutron multiplication factors for cylinders slab or spheres. This relationship is:

Keff = -

Kinf --

I + B 2 g2 with B2 ,2 for spheres, (RTI)2 B2 go2 + ,2 for cylinders, and (NTI) (HT7T)2 B2 ,2 + ,2 + ,2 for slabs !

(777I)2

^

(YT7I)2 (y;7I)2 Kinf, M 2 and A Parameters are. listed in Tables 4-2 and 4-3, and are applicable to homogeneous 00 2 and water mixtures and heterogeneous lattices of UO2 rods in water.

The homogeneous data is taken from ARH-600 and.the heterogeneous values are from DP-1014 and GEBLA i calculations. This latter data in particular has been-conservatively estimated and represents parameters-applicable'to the optimum rod diameter at the given W/F volume ratio.

4.1.9 Infinite Neutron-Multiplication-Factor-(Kinf)

Considerations Ranges for optimum credible mixture densities and heterogeneity (where applicable) have been determined for criticality safety applications by an extensive study of infinite. neutron multiplication factors (Kinf's) for U0 2 and water systems. This study has also been used to infer analytical biases.(as discussed in

Section 4.1.6) ,from a comparison with the corresponding values for the GEMER Monte Carlo Code.

4.1.9.1 Kinf Values-For-Homoaeneous UO+ and Water Mixtures Table 4-4.shows a comparison of Kinf values determined from GEKENO (with Hansen-Roach Cross Sections), GEMER and ARH-600 for homogeneous U02 and water mixtures as a 1

Page 4-7

Reference:

SNM-1097

June 1, 1984 TABLE 4-2 REACTIVITY FORMULA

Kinf, M2 & A PARAMETERS FOR H,OMOGENEOUS U(5.0)O2 & WATER MIXTURES

~

M2 x **

WF H 3 0 Kinf. (cm2 ) (cm) '

O.05 1.23 37.0 8.35 0.10 1.345 31.5 7.45 0.20 1.432 28.5 6.65 '

O.25 1.442 28.0 6.45 0.30 1.440 28.0 6.30 1

0.40 1.410 28.2 6.05 l 0.50 1.346 28.7 5.92 0.60 1.245 29.3 5.86

All Data is extrapolated from ARH-600-

- 12 Inch Water Reflection Page 4-8

Reference:

SNM-1097

June 1, 1984 TABLE 4-3 REACTIVITY FORMULA Kinf, M2 & A PARAMETERS

.FOR HETEROGENEOUS U(5.0)O2 LATTICES IN WATER W/F M2 x**

Volume Ratio Kinf (cm2 ) (cm)

^

1.0 1.452* 32.65* 7.5*

2.0 6.85* t 3.0 6.64* t 4.0 6.52* '

t 5.0 6.44* t 6.0 6.39* t 8.0 1.448* 29.00*** 6.35***

10.0 1.414* 29.00*** 6.30***

12.0 1.374* 29.75*** 6.30*

14.0- 1.326* 29.75* 6.30*

Data Extrapolated From DP-1014, all other is from GEBLA Calculations.

- 12 Inch Water Reflection

- - Estimated Values are from DP-1014 and GEBLA calculations. These latter data in particular have been conservatively estimated and -

represent parameters applicable to the. optimum rod diameter at the given W/F volume ratio.

Page 4-9

Reference:

SNM-1097 l

June 1, 1984 TABLE 4-4 GEKENO, GEMER.& ARH-600 KINF COMPARISON FOR HOMOGENENOUS U(5.0)O 2 + H O 2 MIXTURES u nwn-Equiva- -

Roach lent W/F Uranium Misture M/U Cross niature volume censity Density Atomle Sections GtKEMO Gr.M E R ARH-600 WP H 3 0 Ratio (qs/ce) (qu/ce) Ratio Rinf Minf 2 o Rint t o Kinf

, 0.00 0.00 9.660 10.960 0.00 0.8775 <0.99

0.005 0.06 9.156 10.440 0.15 0.9763 <1.02

0.01 0.11 8.698 9.967 0.30 1.0217 <1.04

t 0.015 0.17 8.279 9.535 0.46- 1.0566 <1.08

t 0.02 0.22 7.895 9.139 0.61 1.1079 <1.10

t t

0.03 0.34 7.215 8.439 0.92 1.1586 1.15 t 0.04 0.46 6.632 7.838 1.25 1.2220 1.19 t 0.05 0.58 6.127 7.316 1.58 . 1.2562 1.$3 t 0.08361 1.00 4.830 5.980 2.73 1.3514 1.32 t-0.10 1.22 4.356 5.491 3.33 1.3779 1.345 t t

0.15 .1.9 3 - 3.293 4.395 5.29 1.4346 1.402 t 0.15432 2.00 3.220 4.320 5.47 1.4420 1.405 t 0.20 2.74 2.583 3.663 7.49 1.4673 1.432 t 0.21490 3.00 2.415 3.490 8.20 1.4678 1.439 t 0.25 3.65 2.076 3.140 9.99 1.4704 1.442 t I

0.26738 4.00 1.932 2.992 10.94 1.4695 1.443 I 0.30 4.70 1.696 2.748 12.84 1.4648 1.440 t 0.31328 5.00 1.610 2.660 13.67 1.4562 1.440 t 0.35 5.90 1.400 2.443 16.13 1.4420 1.430 t 0.35377 6.00 1.300 2.423 16.40 1.4423 1.428 *

. 0.38976 7.00' 1.208 2.245 19.14 1.4275 1.415 t 0.40 7.31 1.163 2.199 19.98 1.4265 1.410 t 0.42194- 8.00 1.073 2.107 21.87 1.4165 1.400 t 0.45 8.97 0.969 1.999 24.52 1.4030 1.385 t 0.45090 9.00 0.966 1'996

. 24.60 1.4026 1.385 t-t 0.47710 10.00 0.878 1.905 27.34 1.3829 1.365 t 0.50 10.96 0.808 1.833 29.96 1.3673 1.346 t 0.5202 11.88 0.750 1.773 32.49 1.3545 1.330 .t 0.55 '13.40 0.671 1.692 36.62 1.3266 1.300 t 0.60 16.44 0.554 1.571 44.94 1.2713 1.245 t' t

0.65 20.35 0.452 1.466 55.65 1.2096 1.175 t 0.67870 23.15 0.400 1.412 63.29 1.1547 1.130 t 0.70 25.57 0.364 1.375 69.91 1.1203 1.095 t 0.75 32.88 0.285 1.294 89.89 1.0219 (0.995* t 0.77450 37.64 0.250 1.258 *102.91 0.9670 <0.940* t

.t 0.00 43.84 0.215 1.222 119.85 0.8974 - t

'O.85 62.11 0.153 1.158 - '169.79 0.7472 - t 0.89720' 95.65 0.100 1.103 261.51 0.5705 - t 0.90 98.64 0.097 1.100 269.66 0.5584 - t 0.95 208.24 0.046 1.048 569.29 -0.3175 - t

Estrapolated .

Page 4-10 Refere'nce: SNM-1097

June 1, 1984 function of the WF H 2O in the mixture. The values under the. column "Hansen-Roach Cross Sections Kinf" are based solely upon the Hansen-Roach Cross Section sets (as determined by direct calculation of Kinf from the fission and absorption cross sections, transfer coefficients, and the fission spectrum). These'are in excellent agreement with the corresponding GEKENO value.

The GEKENO and GEMER. values in this table were calculated using a simple cubical geometry (typically.

200 to 400 cm on a side) and J=O boundary conditions on all six sides. Table 4-4 also. lists equivalent W/F

~

ratios, uranium densities and mixtures densities E (specific gravity) for each of the mixtures based on the-I methods described in Section 4.1.7.

-The Kinf values.in Table 4-4 are plotted in Figure 4-1 and illustrate the following:

4.1.9.1.1 GEKENO with Hansen Roach cross sections is biased '

high relative to GEMER at and around optimum moderation.

This relative bias is estimated to be in excess of 1.0%

based upon a comparison of maximum values.

4.1.9.1.2 ARH-600 Kinf data is biased low relative to GEMER by approximately 0.7% at and around optimum moderation.

4.1.9.1.3 GEMER and GEKENO results indicate spectral differences

- - - - - such that GEKENO overpredicts the Kinf values relative to GEMER by 1.0% at optimum moderation but by even 4 greater amounts at each end of the mixture WF water spectrum. For dry (< 0.01 WF H 2 O) mixtures, this bias is~4 to 5% and for very dilute (> 0.75 WF H 2 O) mixtures, it is 3 to 4%. These regions are of particular interest since they are points at which the UO 2 and water mixtures have Kinfs less than 1.0. (ARH-600 data is not available for these regions and can only be inferred by smooth curve extrapolation).

4.1.9.2 Kinf Values For Heterogeneous UO3 Lattices-In-Water I

As described in Section 4.1.7, heterogeneous square lattices of UO 2 rods in water are characterized by a Water to Fuel Volume Ratio (W/F) given by:

Page 4-11

Reference:

SNM-1097

June 1, 1984 FIGURE 4-1 K-INFINITE FOR U(5)O2 & WATER MIXTURES 4

t

! t t

J 1

. t t

4 t

t

. . ; t t

t t

+,

.I t

t t

t.

t t

t i Page 4-12 1

Reference:

SNM-1097 1

June 1, 1984 4

W/F = (L2_ ,92)

.... 4 wD' 4

For a given W/F Ratio,.then, the Lattice spacing, L, is 4

related,to the rod diameter, D, by l

L=D 7

\ w(1 + W/F)

This relationship has been used to determine lattice dimensions for use in GEMER Monte Carlo calculations of infinite neutron mutliplication factors for heterogeneous UO2 and water. mixtures. The geometry model used in the GEMER calculation consists of a ;

cylindrical rod with the appropriate diameter D centered in a rectangular cuboid with "X" and "Y" dimensions of

L. .The "Z" dimension (with the Z axis also being.the axis of cylindrical rod) was typically taken to be 400 cm . - An infinite lattice was modelled by J=O boundary conditions on all six sides. Material densities were taken to be 10.96 grams /cm3 for the UO 2 in the rod and 1.00 grams /cm 3 for the water moderator outside of the rod.

~

)

Table 4-5 lists the results of these GEMER calculations

for a range of W/F Ratios from 1.0 to 10.0. The rod

) . diameters were selected so as to include the optimum value.for the associated W/F Ratio.

Also shown in Table 4-5 are Einf values calculated with L GEKENO.using Hansen-Roach-Cross Section Sets with.U238 material ids adjusted to correct for heterogeneity of

! the UO 2 in the lattice. The geometry model used in I these calculations was a simple' cube, 200 cm on a side l and reflected via J=0 boundary conditions on all six j sides. The mixture specifications used-were to " smear" l the UO2 and water into a homogeneous mixture with the.

same equivalent W/F~ Ratio-(See Table 4-4) but with a U 23s material ID reduced a number of units sufficient to result in a Kinf at least as great as the corresponding I GEMER value.- The number of units required to be i

subtracted was determined from the-results in Table r

Page 4-13 I

Reference:

SNM-1097 l l

. __ . .- _ _ _ _ .- D

June 1, 1984 TABLE 4-5 GEMER & GEKENO KINF COMPARISONS FOR HETEROGENEOUS UO 2 ROD LATTICES IN WATER Reactivity

. Cell w/r Rod Diameter CERENO GEMER Formula

Volume Ratto (Inches) Kint e Nint t e Nint 1.0 0.00 - t 0.70 - t e

-- t 0.80 0.90 - t 1.00 -- t 1.10 1.452 t 1.20 -- t t

2.0 - 0.00 -- t 0.40 - t; 0.50 1.525 to 0.60 - t 0.70 -- t t

3.0 0.00 -- t 0.20 -- t 0.30 1.540 t 0.40 - t 0.50 -- t

)

4.0- 0.00 -- t l 0.10 -- t 0.20 1.530 t 0.30 - t

. _ _ . - - _ ~ _ . - -

0.40 -- t t

5.0 0.00 -- t 0.10 -- . t 0.20 1.498 t

-- t 0.30 t

6.0 0.00 t 0.10 1.460 t 0.20 t t

8.0 0.00 --

t 0.05 t 0.10 1.448 t t

10.0 0.00 t 0.05 1.414 t 0.10 t

From CEsLA and DP-1014 ,

Page 4-14

Reference:

SNM-1097

June 1, 1984 l

l l

TABLE 4-6 HANSEN-ROACH CROSS-SECTION KINF CALCULATIONS *

(HOMOGENEOUS - U(5.0)O2+HO- 2 MIXTURES)

. E-1MF1 MITES U-234 Mat. 10 W/F=1.0 W/Fe2.0 w/Fe3.0 W/F=4.0 w/F=5.0 W/F=6.0 W/F=0.0 w/F=10.0 92040 ........ ........ ........ ........ ........ ........ ........ ........

92039 ........ ........ ........ ........ ........ ........ ........

92836 ' ........ ........ ........ ........ ........ ........ ........ t 92437 ........ ........ e....... ........ ........ ........ ........

9 92034 ........ ........ ........ ........ ........ ........

t 92835 ........ ........ ........ ........ ........ ........

92834 ........ ........ ........ ........ ........

92e33 ........ ........ ........ ........ ........ t 92832 ........ ........ ........ ........

i 9283t ........ ........ ........ ........

t 92830 ........ ........ ........ ........ '

i 92029 ........ ........ ........

92028 ........ ........ ........ p 92027 ........ ........ ........

92026 ........ ........ I t

9202S ........ ........

92824 ........ ........ i 92623 ........ t 92e22 ........

92e2a ........ t t

92 2. ........

92019 ........ t.

92014 ........ p 92017 ........

92ets ........ t 92sts t 92ste 9 92en3 92st2 t 92ett t 92ste t 92eet 92eos t

92007 t 92006 f

92eos t 92so4 92003 t 92002 '

92e0:

, -t

GE developed based on exact infinite neutron multiplication factor calculations using 16 group Hansen-Roach cross section sets.

Page 4-15

Reference:

SNM-1097

._~ . --- - _ - .

June 1, 1984 1

4-6 which show Kinf values for the . corresponding homogeneous Hansen-Roach Cross Section Sets but with the' specified U238. Material ID. (The values in this table were computed directly from the fission.and absorption cross section values, transfer coefficients and fission spectrum for the specified cross section set).

i .Because the Hansen Roach Cross Sections Sets for heterogeneous 00 and water mixtures were determined as specified above,2the bias associated with GEKENO with these cross section sets is the same as that generally applicable to GEMER, that is, -0.01.

Also included in Table 4-5 are the Kinf values from 8

Table .4-3 which are . used in Reactivity Formula calculations for heterogeneous systems. Since the maximum Kinf in.this data, 1.540, is greater than the corresponding GEMER value, . by more tha'n 0.003 t (which is the magnitude of the assigned GEMER bias as 3, discussed in Section 4.1.6) this bias assigned to the Reactivity Formula with these heterogeneous parameters is 0.00. This only applies to-the optimum case, however, and applications for other conditions will

. require a bias evaluation based on Table 4-6. The range i

of values which may be applicable in such cases is 0.00 -3 to'-0.03. I 1

-4.1.10- Equipment Interaction Criticality safety of interacting equipment, assemblies.

and systems is determined in several ways. These include:

.4.1.10.1 Use of GEMER or GEKENO Monte Carlo Codes for Keff or Kinf; calculations for interacting systems.

4.1.10.2 0s'e of solid angle code (SAC) for equipment interaction l ' evaluation.

4.1.10.3 Use of isolation by' distance (12 feet or the l' orthographic projection, whichever is greater).

4.1'.10.4 Use of neutron absorber panels to isolate v'essels and equipment.

l 4.1.10.5 Disciplined layout of process piping.

l Page 4-16

Reference:

SNM-1097

4 i

June 1, 1984 4.2' UPMP SAFE GEOMETRIES & SAFE BATCH LIMITS Equipment in UPMP used to process or store uranium i compounes is considered to be aeometrically safe if it retains the process material within the dimensional

, - limitatiens of Tables.4-7 or 4-8 under normal and forseeable abnormal process or environmental conditions and provided that it is separated by at least 12 inches

- from concrete ' alls 'or other. significant reflectors.

These tables especially apply to equipment for which the

. assumption of reflection by 12 inches of water is conservative.

For cylindersLand slabs which are not infinite in extent, the dimensional limitations of Tables 4-7 or 4-8 may be increased by means of standard buckling conversion methods.or reactivity formula calculations

~

. incorporating the Kinf, M 2 and A values in the ;

applicable Table 4-2 or.4-3.

i In' addition to geometry control, criticality safety in

< selected UPMP equipment.and operations is based upon uranium mass limits. Normally this mass limit is based upon a safe batch which is defined to be 45% of the minimum critical mass taking into consideration'the geometry of the equipment.and the. enrichment and form of

~

the fissile material. Table 4-9 lists safe batch limits

~ for homogeneous mixtures-of UO2 and water in

, uncontrolled geometries as a function of U235 enrichment -

over the range of 1.1% to 15.04. Safe batch limits for specified geometries are discussed in subsequent sections of this document (especially Sections 4.5 and

[ 4.6).

4.3 CRITICALITY SAFETY .:P-UPMP CYLINDRICAL PROCESSING

'~

L vsssELs l In Table 4-7, the safe geometry cylindrical diameter for ,

l homogeneous U02 and water mixtures is specified to be L

l 9.5 inches. This value is applicable to cylindrical vessels of any. height and constructed of any material .

provided that the material of construction does not reflect neutrons more effectively than a tight fitting 12 inch thick layer of water. Since materials of 1

(

L "9' 'I

Reference:

.SNM-1097

, _ _ - ~ - _ _ _ _ . _ _ . _ _ _ _ _._ _ _. _ ___ _ _ ___ _

l June 1, 1984 TABLE 4-7 SAFE GEOMETRY VALUES FOR HOMOGENEOUS UO 2 -H2 O MIXTURES I

. Nominal Infinite Infinite Weight Cylinders Slabs Spheres Percent Diameter Thickness, Volume, U-235 _ Inches Inches - Liters 2.00 16.7 8.9 105 2.25 14.9 7.9 75.'S 2.50 13.75 7.2 61 2.75 12.9 6.65 51 3.00 12.35 6.25 44 3.25 11.7 5.9 38.5 3.50 11.2 5.6 34 3.75 10.8 5.3 31 4.00 10.5 5.1 29, ;

5.00 '9.5 4.45 24 '

6.00 8.95 4.00 -18.5 7.00 8.45 3.75 17.0 1

For enrichment not specified in this table, smooth curve interpolation may be used.

l

. _ . . TABLE 4-8 SAFE GEOMETRY VALUES

, FOR HETEROGENEOUS MIXTURES OR COMPOUNDSI Nominal Infinite Infinite Weight Cylinders Slabs Spheres Percent Diameter Thickness, Volume, U-235 Inches

Inches- Liters 2.00 11.1 5.6 35.7 2.25 10.5 5.1 30.7 2.50 10.1 4.8 27.3 2.75 9.7 4.6 24.7 3.00 9.4 4.4 22.6 3.25 9.2 4.3 20.9 3.50 9.0 4.2 19.2 3.75 8.9 4.1 18.2 4.00 8.8 4.0 16.9 5.00 8.3 3.6 13.0 6.00 7.9 3.5 11.0 7.00 7.4 --

7.6 1 For enrichment not specified in this table, smooth curve interpolation of values may be used.

Page 4-18.

I

Reference:

SNM-1097 1

"- - -p' r' ,- - .w--art,T-a u--+<--w'- * - ' ~ - - g g - - - - - wr w eyvf- --w -----w 'e r'mw'wgy-Awww-

June 1, 1984 TABLE 4-9 SAFE BATCH LIMITS FOR UO 2 &HO2 (Kgs UO 2)

Nominal Nominal Weight Weight Percent U0 2 UO Percent UO UO U,,, Powder l Peflets 2 U 3 ,4 Powderl Peflets 2 1.1 2629 510 3.4 34.6 31.0 1.2 1391 341 3.6 31.1 28.5 1.3 833 246 3.8 28.3 26.4 1.4 583 193 4.0 25.7 24.7 1.5 404 158 4.2 23.7 22.9 ;

1.6 293.3 135 4.4 21.9 21.4 '

1.7 225.0 116 4.6 20.2 20.0 1.8 183.0 102 4.8 19.1 18.8 1.9 150.6 90.5 5.0 18.1 18.1 2.0 127.5 81.6 5.5 15.4 15.4 2.1 109.2 73.1 6.0 13.8 13.8 2.2 96.8 66.4 7.0 8.3 8.3 2.3 84.3 61.0 8.0 6.9 6.9 2.4 74.7 56.1 9.0 5.9 5.9 2.5 68.9 52.1 10 5.1 5.1 2.6 '

60.5 48.8 11 4.4 4.4 2.7 56.6 45.4 12 3.9 3.9 2.8 52.2 42.9 13 3.5 3.5 2.9 47.6 40.1 14 3.3 3.3 3.0 44.5 38.1 15 3.0 3.0 3.2 38.9 34.1 1 Homogeneous Mixtures 2 Heterogeneous Mixtures NOTE: For enrichments not specified above, smooth curve interpolation of safe batch values may be used.

Page 4-19 l

Reference:

SNM-1097

- - . - - - - . - - . - . - _ _ - . - . . - = _ _ - - - .. - . -

June 1, 1984

, construction can in many cases actually reduce neutron reflection, a criticality safety evaluation has been-performed for selected standard construction materials to evaluate their applicability to UPMP cylindrical

. processing vessels. The conditions or materials considered were:

l' e None (For comparison with the 9.5 inch in diameter in Table 4-7) e Stainless Steel 304 Walls e Carbon Steel Walls e PVC Walls This evaluation consisted of four sets of GEKENO i calculations using JRK Modified Hansen Roach Cross . ;

i Sections for infinite cylinders with standard gauge wall' thicknesses. For processing and criticality safety considerations, only 10 inch cylinders were included and the geometries analyzed were restricted to Schedules 10, 20, 40, and 80. .These correspond to'the dimensions tabulated in Table 4-10. (Schedule 5 and 80S are shown ,

for comparison only). The actual GEKENO geometry model i used consisted of two concentric cylinders with

. dimensions appropriate for the inner diameter and wall

~

thickness, respectively, enclosed in cuboid water reflector whose X and Y dimensions were at least 24 inches greater than 10.75 inches. -The axial (i.e. "Z")

dimensions of the concentric cylinders and cuboid were all 200 cm and J=0 bounding conditions were applied at~

. the top and bottom. Mixture and material specifications

were taken to be the following: .

e Fuel - Full density homogeneous 002 and water mixtures as defined in Section 4.1.7.

i e Stainless Steel 304 - KENO /Hansen Roach Cross Section 4

Set No. 200 with a (GEKENO Input) density of 1.00.

e Carbon Steel' ' KENO /Hansen Roach Cross Section Set j No. 100 with a (GEKENO Input) density of 1.00.

j Page 4-20

Reference:

SNM-1097

_. ~ _ . _ _ _ _ . . _ . . - _ _ . _ _ _ _ _ _ . _ . . . - _

O June 1, 1984 TABLE 4-10 DIMENSIONS OF 10" DIAMETER, SEAMLESS & WELDED PIPE

Outside Diameter Thickness Inside Diameter Schedule (Inches)- -

-(Inches)- -(Inches) 5 10.75 0.134 10.482-

- 10 10.75 0.165 10.420 '

20 10.75 0.250 10.250 40 10.75 0.365 10.020 80 10.75 0.593 9.564 80S _

10.75 0.500 9.750

-*From ARH-600, Page II.A.2-6.

i Page 4-21

Reference:

SNM-1097 j

- i June 1, 1984 e PVC - Materials and densities specified in Table II.

F.1 of ARH-600 (Hydrogen = Hansen-Roach Cross Section Set No. 1102) e Water. Reflector - KENO /Hansen Roach Cross Section Set No. 502 with a (GEKENO Input) density of.1.00.

The results of the GEKENO calculations for the four cases are shown in Tables 4-11 through 4-14. These can.

l be sum'marized as follows:

4.3.1 - The maximum Keff i a calculated with GEKENO for a 9.5 inch bare infinite cylinder (which is a safe geometry as specified in Table 4-7) is only This' .'

is an indication of the conservatism inherent in the use 4

of GEKENO and Hansen Roach Cross Section Sets in this application.

4.3.2 For the three materials of construction considered in- i

the analysis, the geometrically safe cylindrical

1. diameter is significantly greater than the 9.5 inch Safe Geometry Value. These dimensions are:

e For Stainless Steel 304, Inch Schedule t inch ID) '

For Carbon Steel, Inch Schedule inch t ID)

For PVC, Inch Schedule inch ID) 4.4 UPMP FIXED NEUTRON ABSORBER PANELS l Preliminary UPMP process design reviews _have stressed the need for geometry control in criticality safety in e addition to the use of historically-acceptable density or mass controls. Therefore, the UPMP design utilizes neutron absorber panels as a form of geometry control to be employed in addition to density and mass controls.

UPMP fixed neutron absorber panels are fixed in a t matrix of encased in t The absorber panels are affixed to and part of the !

. chemical vessel structures. Inasmuch as their typical l

weights range.from lbs, they are not t !

subject to movement or removal. '

Page 4-22

Reference:

SNM-1097

i June 1, 1984 TABLE 4-11 GEKENO Keff VALUES FOR BARE INFINITE CYLINDERS (Homogeneous U(5.0)O 2 + H2 O Mixtures)

Infinite Cylinders WF H,0 Keff a 9.0" in Diameter 0.1 t 0.2 t 0.25 t 0.3 t 0.4 t 0.5 t i,

-9.5" in Diameter 0.1 t 0.2 t 0.25 t 0.3 t 0.4 t 0.5 t 10.0" in Diameter

~

0.1 t-0.2 t 0.25 t 0.3 t 0.4 t 0.5 t 10.5" in Diameter 0.1 t 0.2 t 0.25 t 0.3 t 0.4 t 0.5 t Page 4-23

Reference:

SNM-1097

l June 1, 1984 l

TABLE 4-12 GEKENO Keff VALUES FOR SS-304 INFINITE CYLINDERS (Homogeneous U(5.0)O 2 + H2 O Mixtures)

Infinite Cylinders WF H,0 Keff i o 10.0" Schedule 0.1 0.2 :

0.25 t 0.3 t 0.4 t 0.5 t 10" Schedule 0.1 t 0.2 t 0.25 t 0.3 t 0.4 't 0.5 .1

-10" Schedule 0.1 ^

0.2 O.25 t 0.3 t 0.4 t 0.5 t

-10" Schedule 0.1 t 0.2 t 0.25 t 0.3. t 0.4 t

. 0.5 t Page 4-24

Reference:

SNM-1097

June 1, 1984 i

TABLE 4-13.

GEKENO Keff VALUES FOR CARBON STEEL INFINITE CYLINDERS (Homogeneous U(5.0)O 2 + H2 O Mixtures)

Infinite Cylinders WF H,0 Keff i a 10" Schedule 0.1 t 0.2 t

' 0.25 t 0.3 t 0.4 t 0.5 t I.

' 10" Schedule 0.1 t 4

0.2 t 0.25 t 0.3 t 0.4 t 0.5 t 10" Schedule 0.1 t 0.2 t 0.25 t 0.3 t 0.4 t 0.5 t 10" Schedule 0.1 t 0.2 _t 0.25 t 0.3 ~t 0.4 t

. 0.5 t 1

'f i Page 4-25

Reference:

SNM-1097

June 1, 1984 l TABLE 4-14 GEKENO Kef f VALUES FOR PVC INFINITE CYLINDERS (Homogeneous U(5.0)O 2 + H2O Mixtures)

Infinite Cylinders WF H,0 Keff i o 10" Schedule 0.1 t 0.2 0.25 .- !

0.3 t 0.4 t 0.5 t 10" Schedule 0.1 '

t 0.2 t 0.25 t 0.3 ,

t 0.4 t 0.5 +;

I 10" Schedule 0.1 t 0.2

0.25 ~

0.3 t 0.4 t 0.5 t 10" Schedule 0.1 t 0.2 t 0.25 t 0.3 t 0.4 t 0.5 t Page 4-26

Reference:

SNM-1097

I June 1, 1984 )

l l

An extensive quality assurance plan is employed to certify the proper fabrication of the panels, to verify the presence of neutron absorbers in them, and to assure their proper installation.

was selected as the matrix material t because of ease of fabrication and desirable nuclear properties Since the neutron absorber panels are encased in they have adequate fire resistance t ,

against credible fire hazards in the facility.

Continued effectiveness of the neutron absorber panels is assured via the tank designs, t 4.5 CRITICALITY SAFETY OF UPMP PROCESSING VESSELS Because of large throughput volumes and high flow rates',

scrap and waste processing in WMD fuel manufacturing operations require large process and storage vessels.

In the Uranium Process Management Project, this need has been satisfied while at the'same time complying with the requirement. for criticality safety by geometry control through the use of large diameter tanks with t neutron absorber panels. t Figure 4-2 shows a typical example. There are such t vessels in the UPMP facility; of these, the t dissolvers, are designed for heterogeneous t UO 2 and water mixtures. The remaining t vessels are designed for homogeneous UO 2 and water mixtures as based upon the following criticality safety analysis.

4.5.1 Vessel Descriptions UPMP operations with processing vessels are t described in detail in Sections 2.1 through 2.4 and 3.1. .

In general, these vessels are inch thick t tanks with of inches and t heights of . feet. The vessels are shielded t neutronically by t neutron absorber panels which are t Page 4-27

Reference:

SNM-1097

June 1, 1984 FIGURE 4-2 UPMP VESSEL DESIGN t t

t t

o' t

t t

j t

- ; t -

i t

t t

t ,

t q.

1

! .t t

t t

. t t

t ,

t t >

t i

! t t i i

t t

t ,

l i

A Page 4-28

Reference:

SNM-1097 t- wrw. .,,,..--WT " - * **N7"C--W""7--- W'W' -77 +mW--F"-N"1 3"TT""""Y-* -W'"'- mPw+ m-r twt-@~ -- --? w--wa-'+ ww ww-wer% m---- e--w.a w-w -wr*1e-

June 1, 1984 inches thick'and about inches' longer than the t tanks. The neutron absorber panels are t spaced from . inches away from the tank t and are designed both to maximize the t of the tank t and minimize the neutron interaction with nearby t equipment t Tables 4-15 and 4-16 present a detailed description of the vessels and their associated neutron t

. absorber panels for each of the vessels. t 4.5.2 Neutron Absorber Panels

, The vessel neutron absorber panels have been t designed to be stable fixed in a high t density matrix. Table 4-17 lists the t 4

specifications.for this material. Each panel is completely encased in gauge t i, t 4.5.3 Analytical Methods The effectiveness of the neutron t absorber panels described in the previous section has been analyzed using the GEKENO Monte Carlo Code with JRK modified Hansen-Roach cross section sets.

This analysis consisted of sets of calculations of the neutron multiplication factors for representative geometries as illustrated in Figure 4-3 with three different t for the tanks and neutron absorber t panels. The of the tanks were. t inches and were selected to conservatively span the t range of geometries specified in Tables 4-15 and 4-16.

Figures 4-4, 4-5 and 4-6 show the detailed dimensions of the vessels that were used in the geometry t i

models. Reflection was modelled on the top and bottom

~

of the tank by a tight-fitting 12 inch water reflector i as shown in Figure 4-7.

Materials of construction and fuel mixtures were also l modelled conservatively. These were taken to be:

Fuel: Full density homogeneous 00 2 and water mixtures l (as discussed in Section 4.1.7).

Steel (GEKENO mixture No. 200 with I a " mixture density" of 1.0).

L l

l Page 4-29 l-

Reference:

SNM-1097 i

. , ~. , .

t

,Y ,

( -

T June 1, 1984

. n ,

s

. . , .^

~ .

%w y #

p 7

/

TABLE 4-15' '.

< VESSEL, DESCRIPTIONS t

. . ' .~

> e _'

~ ; /

Neutron Absorber

/ ~ ,

Reaght Vess 1 Coometryl

" /). Volume Fanel t

vessel ID vessel Description (In) , (Ca11ons) t

. i 1

1. Tank .

500- IA IIA t

'r f . ", / .

t

2. Tank s ss

'- 500- IA IIA t

,. - t

, '~,

3. Tank 500 IA !!A

7 600 IB IIB e

4. Tank

/ t

5. Tank e l' '

600 IB !!B t t

6.' Tank 600 IB Its t 1

/*

7. 600' IB !!B t Tank .

/ ,

t B. Tank .

~

600 IB ; IIB t

. t

9. Tank 500 IA IIA t t

10. Tank 500 IA IIA t

t

600 IB IIB t

11. Tank t

-' ~>

12. Tank .

< 600 18 115 t T

13. Tank ,

600 IB !!B 1

.. I Tank 500 IA IIA t

14. ,

a e . t Tank

.-,* 500 1A IIA t

15. .? ?

16. Tank 600 j IB IIB

17. Tank 600 IB IIB t 9-

15. Tank

750 1C IIC tl

ti

19. Tank 750 IC Itc' tj

t-

20. Tank _

600 IB 115 t

21. Tank ,

, [*

600 It !!B t t

22. Tank .

d 750 IC !!C t

t

23. Tank ^

. 750 IC !!C t

~ t 24 Tank %.

500 _ IA ', !!A t

' t

,s ,e

25. Column * - 375 IB . 115 t j

26. Columri -. .. 375 15 IIB t t

27 y 375 IB 11B t

. Column iss-204 t. att,. 3/B Inen thick wa11e eae,st a. +oted ,

8se. Tant. 4-i6 '

85/4 inch thick walls , , .

,,7 3

[

J g 4 i ~ . Page 4-30

Reference:

INM-1097 1

- . . _ , _ ,- - - r

June 1, 1984 TABLE 4-16 ,

GEOMETRY DESCRIPTION OF NEUTRON ABSORBER PANEL neight Number t Destenation Type (inches) (Inches) (Inches) of Panels I.A t 8 t C , t D t E t i

!!.A t a t c t t

D t E t TABLE 4-17 NEUTRON ABSORBER PANEL COMPOSITION A. Minimum Panel Density t B. Chemical Constituents Composition Range (%)

t t

t C. Elemental Constituents Element - Atomic Mass Composition Range (%)

t t

D. Densities Of Constituents Density Range (grams /cm3 )

t t

, Page 4-31

Reference:

SNM-1097

- , .et

' t.

U, ,, -

I June 1, 1984 s il j FIGURE 4-3 ETORAGE' TANK GEOMETRY MODEL t

.q ,

.. .- -)

i

't r f

i i~

t

/~'e ! t t

a .

-+ .

, t s.'

t i' ' '

s t

t

, . ; t i

9

s. s N,. t

. t t t t

,. +,

's i t

. . t

.. . T ,' +

L t

~

< s t r '

.t

. . t

>. t t

t t

, t .,

t

! T',

9

. t t

t i

/ .

1 P i .

, s i.

- i '. Page 4-32

Reference:

SNM-1097 O

} ..

4 June 1, 1984 FIGURE 4-4 VESSEL MODEL -t i

I t

t t

, t t

t 3, .,

' t t

t

> - t t

t t

k t

!!' t t

t t

e> t t

t-t

.i Page 4-33

Reference:

SNM-1097

June 1, 1984 FIGURE-4-5 VESSEL MODEL t l

^

t .

t j t

t t

t

t t

t t

t-t

+1 t

t

+

t t

t Page 4-34

Reference:

SNM-1097

June 1, 1984 FIGURE 4-6 VESSEL MODEL t t

t t

t

. . ; t t

t t

9 t

t t

t Page 4-35

Reference:

SNM-1097

June 1, 1984 FIGURE 4-7 MODEL t t

t t

t s'

t t

- ; t t

t t

1 1

t t

+

t t

t Page 4-36

Reference:

SNM-1097

June 1, 1984 Mixture specification in Table t 4-18.

Water reflector: . Full density water (GEKENO Mixture No.

502 with a " mixture density" of 1.0).

The calculations performed in this analysis were the-following:

4.5.3.1 For each of the vessel models, the t effective neutron multiplication factors for single water reflected vessels with " normal case" fuel 4

densities of 100, 250, 400 and 750 gm U/1.

4.5.3.2 -For each of the vessel models, the t effective neutron multiplication factors for single water reflected vessels with fuel mixtures ranging from 0.10 to 0.50 weight fraction of' water. (This range ,

includes optimum moderation and establishes that the '

vessels are geometrically safe).

4.5'3.3 . The effective neutron multiplication factor for a

" damaged" worst case vessel which has lost its t '

neutron absorber panels but which is still in' conformance with~its density control (assumed to be t gm U/1). The worst case vessel t determined by 2 above was found to be the inch t model, however, dif ferences between .models were t not significant.

4.5.3.4 The effective neutron multiplication factors for the damaged inch geometry model with fixed masses. t of uranium (100 Kg, 250 Kg, 500 Kg and 1,000 Kg) and

.with varying weight fractions of water. The geometry model in this case assumed that the fuel mixture filled 4

the tank ~from the bottom up to a height equivalent to

.the U mass content and the rest of the tank is. filled with water. These results permit the determination of the-minimum critical mass for the vessels under accident conditions in which the neutron absorber panels are lost.

~

4.S.3.5 The infinite neutron multiplication factor for a planer array of undamaged inch geometry model vessels t with its most reactive fuel mixture and with a 16 inch concrete reflector on the bottom. (The top reflector is Page 4-37 Referencer. SNM-1097

Y s

June _1, 1984 TABLE 4-18 MIXTURE SPECIFICATIONS t FOR GEKENO t

i' t

t t

t

. ; t t

t t

't 1

Page 4-38

Reference:

SNM-1097

p .

4:

1 June 1, 1984-the same 12 inch water reflector). The J=O reflector boundary condition was placed at the t neutron absorber panel' surface and t the region outside of the panel was assumed to contain

various levels of interspersed water. These calculations thus establish that interaction between vessels (by virtue of the neutron t absorber panels) does not lead to a critically unsafe condition.

4.5.3.6 The effective neutron multiplication factor for a single undamaged inch geometry model with its most t

. reactive fuel mixture and with various levels of interspersed water in the air gaps between the tank and neutron absorber panels. These t calculations establish criticality safety of credible upsets that could lead to such conditions..

4.5.4 Results ;

The results of the calculations described in the previous section are listed in Tables 4-19 through 4-23.

These are summarized as follows:

4.5.4.1 The maximum neutron multiplication factor for normal case operations with undamaged vessel (i.e., neutron absorber panels intact) and density control (of no more than gm U/1 as discussed in Chapter 2) is t This value is in conformance with the 0.90 t limit specified in Section 4.1.5.1. The data in Table 4-19 also shows no significant dependence on the geometry model- t 4.5.4.2 The maximum neutron multiplication factor for optimum moderation fuel mixtures in undamaged vessels is

! and occurs for a 0.25 WF H O fuel t mixture in the inch geometrymode$' case. The t value is within the 0.97 limit specified in Section 4.1.5.2 and consequently the process vesssels t

, are geometrically safe for homogeneous U0 2 and water mixtures. This has been verified by a calculation performed with.the GEMER Monte Carlo code using the 52 inch geometry model and the same materials and t densities as for the WP H 2O = 0.25 case. The GEMER result is a Reff of is compared to the t GEKENO value of in Table 4-20. t Page 4-39

Reference:

SNM-1097

l l

l June 1, 1984 i

1 TABLE 4-19 VESSEL CALCULATION RESULTS t FOR NORMAL CONDITIONS Geometry Model Fuel Mixture U GEKENO

, (Inches) Density (grams / liter) Keff 2 o 100 t-250 t 400 t 750 -

3, t t

100 t 250 t 400 t 750 t 100 1 250 t 400 t 750 '

Page 4-40 Reference SNM-1097

L June 1,.1984 l

I TABLE 4-20 VESSEL CALCULATION RESULTS t FOR OPTIMALLY MODERATED FUEL MIXTURES Geometry Model Fuel GEKENO (Inches) WF H,0 Keff 2 o t 0.10 t 0.20 t 0.25 t 0.30 i t

0.40 t 0.50 t 0.10 t 0.20 t 0.25 t 0.30 t 0.40 t 0.50 t 0.10 t 0.20 t 0.25 t 0.30 t 0.40 t 0.50 t I

i Page 4-41 Reference SNM-1097

l l

[ June 1, 1984 TABLE 4-21 VESSEL CALCULATION RESULTS t FOR ACCIDENTS INVOLVING LOSS OF NEUTRON ABSORBER PANELS Geometry Fuel Model Mixture U Density, GEKENO t inches (grams / Liter) Keff 2 o A. DENSITY CONTROL RESULTS 250 t Fuel U Mass ;

WF H,0 Kg '

B. MASS CONTROL RESULTS 0.25 100 t 250 t 500 t

~1000 +1 0.30 100 t 250 t 500 '

1000 0.40 100 t 250 t 500 t 1000 t 0.50 100 t 250 t 500 t 1000 t O.60 100 t 250 t 500 t 1000 t Page 4-42 References SNM-1097

M June 1, 1984 FIGURE 4-8 TANK CRITICAL MASS VARIATION t WITH H 2O WEIGHT FRACTION i

t t

t

, t

- ; t t

t t

t

. t l t t

.t t

t t

. t

t t

, t t

t t

t i

)

Page 4-43

Reference:

SNM-1097

June 1, 1984 l

l I

TABLE 4-22 VESSEL CALCULATION RESULTS t FOR INFINITE PLANAR ARRAYS Geometry Model Fuel Interstitial GEKENO (Inches) WF H,0 H,0 Kinf 1 o t 25 0.00- 1 8

0.01 t 0.05 t 0.10 t TABLE 4-23 VESSEL CALCULATION REJULTS t FOR ACCIDENTS RESULTING IN INTERSPERSED WATER IN AIR GAPS Panel Gap Geometry Model Fuel H 2 O Density GEKENO (Inches) WF H,0 (qm/cm 3) Keff 2 o t 0.25 0.00 t 0.05 t 0.10 t 0.20 t 0.50 t 1.00 t Page 4-44

Reference:

SNM-1097

> +

June 1, 1984 4.5.4.3 As in the results for the normal case, no significant dependence on geometry model is evidenced by the t optimally moderated data. Accident condition cases considered in the remaining part of the analysis have utilized only the inch (or in some cases, t f

inch) model. t 4.5.4.4 The effective neutron multiplication factors shown in Table 4-21 for the damaged inch vessel t can be extrapolated to yield a conservative estimate of the minimum critical mass in the vessel under the assumption that the neutron absorber panels have failed.

Using a criteria of Keff + 3a = 0.95, these extrapolated values have been plotted in Figure 4-8 and indicate a minimum critical mass of kg of uranium. The 45% t safe batch value is thus kg of uranium. t 4.5.4.5 The neutron multiplication factor for an infinite planar s

array of undamaged inch vessels with no ; t interstitial water is 0.9578 1 0.0029. This is an increase of about 2% over the Keff value for a fully reflected tank and signifies negligible interaction between tanks especially when compared to the Kinf of 1.4722 1 0.0026 for a homogeneous 002 + 0.25 WF H O 2 4

mixture. The results in Table 4.22 also indicate a slight spectral effect for gms/cm 3 of t

~

interstitial water. In this case, the Kinf + 3a value is still less than.1.0.

. 4.5.4.6 Table 4-23 shows that a single undamaged process vessel

. with an optimally moderated fuel mixture remains critically safe for interspersed water in the t up to gm/cm3 . Since the gaps between the t vessels and neutron absorber panels are t water in excess of this level is t incredible. In addition, the fixtures provided for holding the neutron absorber panels in place also l prevent interspersed water in the gaps.

In summary, the calculations described in the above sections have established that the processing t vessels with neutron absorber panels comply with the requirements listed in Section 4.1.5 for both normal and i specified accident conditions and hence are.

geometrically safe. The specified accident conditions considered were loss of neutron absorber panels and i

I Page 4-45

Reference:

SNM-1097

June 1, 1984 presence of interspersed water in the air gap between ~

the tank and neutron absorber panels. Loss of t geometry (other than the neutron absorber panels) has

. -not been discussed in this section (it is addressed in

the applicable parts of Chapters 2 and 3) . because .

criticality safety in such cases is based upon density or mass controls.

4.6 CRITICALITY SAFETY OF UPMP DISSOLVER & LEACHING VESSELS UPMP scrap processing operations employ different types of dissolving vessels depending on the nature of t the material being processed. Oxidized high grade scrap (U02 hard scrap, powder or high grade sludge) is processed in kg batches in dissolvers. Low grade scrap (predominantly incinerator t ash and oxidized low grade sludges) is processed in inch Schedule cylindrical leachers. t Low grade' scrap leaching operations are limited to . ; t individual batches of one three or five gallon pail at d time with a uranium content of less than one safe batch (18.1 kg 002). Because dissolver and leacher operations can potentially involve oxidized 'hard . crap, t these vessels have been designed to be geometrically t safe for heterogeneous lattices of U0 2 in water. These

. heterogeneous lattices have been very conservatively }

t assumed to be square lattices of (infinite) rods in water with optimum rod diameters and water to fuel (W/F) volume ratios. The applicable criticality safety l analyses are contained in the following two sections.

4.6.1 Dissolvers The UPMP dissolvers are large t vessels with neutron absorber t panels similiar to the design shown in Figure 4-2. The design and form of the neutron absorber panels are the same as those in the processing vessels (as t specified in Table 4-17) and comply with the same quality. and operational considerations as discussed in Section 4.4. The key differences between the t processing' vessels and dissolvers are the t

, dissolver geometry, analytical methods and material

contents.

1 1

Page 4-46 i

References SNM-1097

June 1, 1984 4.6.1.1 Dissolver Geometry Table 4-24 describes the important geometrical features of the dissolvers. The significant difference t from the geometry for processing vessels is t that the latter have an t inches as compared to inches for the t dissolvers.

4.6.1.2 Analytical Methods 4.6.1.2.1 Geometry Model Figures 4-9, 4-10, and 4-11 show the geometry model used in this criticality safety analysis of the t dissolvers. As in the case of the criticality safety analysis of processing vessels, key geometric t dimensions have been conservatively adjusted to reflect fabrication tolerances and, where applicable, corrosioni, allowances.

1 4.6.1.2.2 Fuel Mixtures dissolvers will be used in operations with t oxidized sludge, powder, and pellets predominantly in the form of U 308 powder. Since the potential exists for heterogeneity due to particle size effects, the t dissolvers have been analyzed for the case of optimum lattices of UO 2 rods in water. (This is overly conservative and is subject to future revision based on particle size studies in oxidized scrap.) Since modelling of rod lattices in vessels is not t practicable with the current version of the GEMER Monte Carlo Code, this analysis has been performed with GEKENO using the heterogeneous cross section sets discussed in Section 4.1.9.2. All other material specifications are the same as those described in Section 4.5.3 for (homogeneous) processing vessels. t Although the GE1ER Monte Carlo Code is not directly applicable to tne dissolver criticality safety t analysis, it has been used to validate GEKENO with the heterogeneous Hansen-Roach cross section sets. This has been done by performing comparison calculations with GEKENO and GEMER for infinite planar slabs of U0 2 rod lattices in water with metal thicknesses, air gaps and neutron absorber panels t l Page 4-47 References SNM-1097 l

June 1, 1984 TABLE 4-24 DISSOLVER DESCRIPTION t Vessel Number: t Vessel Height: Inches s' Inches t Inches t vessel Wall Thickess: Inches t Vessel Volume Gallons !

Neutron Absorber Panel Types *

see Table 4-16 l

Page 4 48 i

Reference:

SNM-Iog7

I I

i June 1, 1984 FIGURE 4-9 STORAGE TANK GEOMETRY MODEL t L

t

! t

, t

!' . t t

t t

t j t t

t t

t l t t

t t

I t-l t l t

[ t t

t t

t t I

t :

I t t

t t

l l

\

l i

Page 4-49 Reference SNM-1097 l

June 1, 1984 FIGURE 4-10 VESSEL MODEL t FOR DISSOLVER t i

t t

3, t t

t t

't t

t t

t Page 4-50 References SNM-1097

June 1, 1984 FIGURE 4-11

GEOMETRY MODEL t FOR DISSOLVER t i-

+

t

't t

J t

t

- i t i t t

t t

i t

t t

t t-

t t

t

.t t

t t

i t t

t t

i 4

l Page 4-51 Reference SNM-1097

June 1, 1984 sides) representative of the dissolver i geometry. The GEMER calculations were performed for a

)

lattice of 0.35 inch in diameter UO rods in water with '

W/Fsvolume ratio of 3.0 and the GEKbNO calculations were l performed using the heterogeneous U0 2 and water cross section set for the same W/F volume ratio. The t

, volume ratio was selected for these comparisons t because it is closest to optimum. Table 4-25 contains the results of this comparison study and indicates no significant bias of GEKENO relative to GEMER.

~

4.6.1.2.3 Description of Dissolver Calculations The calculations performed in this analysis were the

following:

4.6.1.2.3.1 The effective neutron multiplication factor for a single water reflected dissolver,,with fuel- t water mixtures ranging from water to fuel volume ratios;,

of 1.0 to 6.0.- This range spans optimum moderation 4

conditions and establishes that the t dissolvers are geometrically safe for heterogeneous UO 2 4

in water.

4.6.1.2.3.2 The effective neutron multiplication factor for a i single dissolver which has lost its neutron 3 absorber panels and which contains uranium masses of 4 100'kg, 250 kg, 500 kg, and 1,000 kg. The results of these calculations permit the determination of the minimum critical mass for the dissolvers under accident conditions which compromise the integrity of the neutron absorber panels.

4.6.1.2.3.3 The infinite neutron multiplication factor for a

~

planar array of undamaged dissolvers with the t most reactive (heterogeneous) fuel mixture and with a 16 inch concrete reflector on the bottom. (The top reflector is the same 12 inch water reflector as for the two previous cases). The J=0 reflector boundary condition was placed at the neutron- t j absorber panel surface and the t region outside-of the panel was assumed to have various levels of interspersed water. This set of calculations was' performed to establish that interaction between dissolvers and other processing t vessels does not lead to a critically unsafe condition.

4 Page 4-52

Reference:

SNM-1097 1

June 1, 1984 TABLE 4-25 VALIDATION OF GEKENO HETEROGENEOUS CROSS SECTION SETS t

l l

Slab

W/F Thickness GEKENO GEMER

Volume Ratio (Inches) Keff i o Keff i o 2.0 4.3 t 5.0 t 5.6 t

0.35 Inch in Diameter U(5.0)O 2 Rods Page 4-53

Reference:

SNM-1097

. . - _ - . . -. . . = . - - - _ _ - .- - - - _ . - _ - _ -

June 1,1984 l

4.6.1.2.3.4 The effective neutron multiplication factor for a-single undamaged dissolver with its most t reactive (heterogeneous) fuel mixture and with various levels of interspersed water in the air gaps between the tank and neutron absorber panels. t 4.6.1.3 Results The results of the calculations described in the '

. previous section are listed in Tables 4-26 through

4-29. These are summarized as follows:

4.6.1.3.1 dissolvers with neutron absorber panels are geometrically safe for heterogeneous 004 mixtures, with "

a maximum GEKENO. result (in Table 4-26)'of t t

4.6.1.3.2 Figure 4-12 shows a plot of the-results in Table 4-27 extrapolated for uranium masses corresponding to a Keff;,

+ 3 o = 0.95 value. From this figure, the minimum critical.(heterogeneous) mass in the t dissolvers under accident conditions in which the neutron absorber panels are compromised is in excess of kg of uranium. -t

4.6.1.3.3 As shown in Table 4-28, the maximum GEKENO neutron i multiplication factor for an infinite planar array of

~~ ~

~dissolvers with the most reactive t

! (heterogeneous) fuel mixture is ^

Infinite planar arrays are thus critically safe. In addition, this result, along with those in .1 and .2 above are less than'the corresponding values for processing vessels and hence the t

. dissolvers can be replaced by processing t

, vessels in UPMP facility interaction analyses.

4.6.1.3.4 The results in Table 4-29 show that the Keff for dissolvers with full density water in the air t gaps between the tank and neutron absorber t panels is only increased by t and the dissolvers are still t criticalty safe under this condition.

In summary, dissolvers are geometrically safe t vessels (for UO 2 rod lattices in water) and are less l

i

Page 4-54

Reference:

SNM-1097

. ,- , ,---,,-,,,r ,.,,.,--,-m,-,-------,-,,,---,,,,,.

,,.---,,.--y-,_,,-,_.-,,.n_,,_

4 June 1, 1984 .

TABLE 4-26 DISSOLVER CALCULATION RESULTS t FOR OPTIMALLY MODERATED FUEL MIXTURES W/F GEKENO Volume Ratio Keff i o 1.0 t 2.0 t 3.0 t ,

4.0' t 5.0 t 6.0 t TABLE 4-27 DISSOLVER CALCULATION RESULTS t FOR LOSS OF NEUTRON ABSORBER PANEL CONDITIONS WITH MASS CONTROL W/F U Mass GEKENO Volume Ratio Kg Keff i a 3.0 100 t 250 t 500 t 1000 t

'4.0 100 t 250 t 500 t 1000 t 5.0 100 t 250 t 500 t 1000 t 6.0 100 t 250 t ,

500 t 1000 t !

1 Page 4-55

Reference:

SNM-1097

June 1, 1984 TABLE 4-28 DISSOLVER CALCULATION RESULTS t FOR INFINITE PLANAR ARRAYS W/F Interstitial GEKENO Volume Ratio H,0 Reff i o 3.0 0.00 t 0.01 0.05 .'

O.10 t i.

TABLE 4-29 y DISSOLVER CALCULATION RESULTS FOR ACCIDENTS RESULTING IN INTERSPERSED WATER IN AIR GAPS Panel Gap W/F H 2 O Density GEKENO Volume Ratio (qm/cm 3) Keff i e 3.0 0.00 t 0.05 t 0.10 t 0.20 t 0.50 t 1.00 t i

i Page 4-56 Reference SNM-1097

June 1, 1984 FIGURE 4-12 DISSOLVER TANK t CRITICAL URANIUM MASS VARIATION WITH W/F t

t t

t

', t t

t t

, t t

t t

t Page 4-57 References SNM-1097

~ j

~ o ,

' "' # ^

June 1,~1984

~

reactive than -

processing vessels under .

similiar~

- y, conditions.+

4.6.2 Leachers A comparison of Tables 4-4, 4-5, and '4-12 indicates that inch Schede 21e Stainless Steel 304 infinite t cylinders are geometrically safe for heterogeneous U0 2 rod. lattices in water. This is demonstrated in the following sections and is the basis for the design of the; leachers in the UPMP scrap processing t f operations.

I 4.6.2.1 , Geometry Description '#

Table 4-30 shows geo$htry specifications for optisum UO 2 rod lattices in water and inch Schedule t cylinders. +

l

' o i, 4.6.2.2 Analytica1' Methods '" <

. t This criticality safety analysis has. been performed with l the GEMER and GEKENO Monte Carlo Codes and the i-i reactivity formula with the heterogeneous mixture ,

parameters.in Table 4-3. The GEMER calculations were t performed for cases with W/F volume ratios of 1.0 to 6.0 3 with explicit modelling of rod lattices in the leacher.

Figure 4-13 shows a plot of the GEMER Geometry Model for the W/r volume ratio = 3.0 case. GEKENO calculations were performed for W/F volume ratios of 1.0 to 10.0 using the heterogeneous Hansen-Roach Cross sections discussed Section 4.1.9.2. The Reactivity' Formula calculations were performed for geometrically unrestricted safe batches (18.1 kg 00 ). Full water l 3 reflection was assu.ned !n all cases. 2 &

4.6.2.3 Results ',

Table 4-31 shows the resulte of these calcui$tions. The maximum neutron multiplication factor' calculated with GEMER is

and occurs for'a W/F volume t ratio of 3.0. Since the radius of the " inch t

, schedule ~ cylinder it cm,4the safe batch t results with the reactivity formula and the Jef f values from.GEKENO imply that under safe batch controls, the Reff of the 10 inch Schedule stainless steel 304 t i

9 Page 4-58 References SNM- 10'J 7 '

g

June 1, 1984 TABLE 4-30 GEOMETRY SPECIFICATIONS FOR SCHEDULE INFINITE CYLINDERS t A. Rod Lattice Parameters:

Optimum W/F Rod Diameter

Lattice Spacing Volume Ratio (Inches) (Inches) 1.0 1.10 1.378646

. 2.0 0.50 0.767495 . ;

3.0 0.30 0.531736 '

4.0 0.20 0.396333 5.0 0.20 0.434161 6.0 0.20 0.468947 B. Inch Schedule Cylinder Parameters: t OD Inches t Wall Thickness Inches t ID Inches (Radius = cm) t

Reference Table 4-5 Page 4-59 References SNM-1097

}

} /

June 1, 1984

FIGURE 4-13 p GEMER GEOMETRY MODEL

) FOR LEACHERS ,

,r. t f

V , t t

t t

, 0 t

i t

' t s t i, t.

4. t 4 . t

' -? t w 4 t i

\ f n g 4

s t

's t q , .

,. 4 I

,i t t

t

, t -(

t-4

't ,

i t

t t

t

-t t

t

, t

. t 1

s.

p 5

.I i 4 Page 4160

~

Reference:

' SNM-1097 ,

3 'l

June 1, 1984 TABLE.4-31 SCHEDULE LEACHER t CALCULATIONAL RESULTS

+-

A. Infinite Cylinder Results:

W/F GEMER GEKENO Volume Ratio Keff a Keff i o

'l 1.0 t 2.0 t s

3.0 t 4.0 t 5.0 t i+ . , ,

6.0 . ;

t

. ,. ; 8.0 t

, 10.0 t n

, B. Safe Batch Results:

W/F Safe Batch

Reactivity Formula Volume Ratio Radius (cm) Keff 1.0 0.675 t 2.0 t-3.0 t

i. 4.0 t 5.0 t 6.0 t 8.0 0.898 t' 10.0 0.906 t 12.0 0.898 t 14.0 0.888 t

18.1 Kg U(5.0)O 2 4 .

Page 4-61

Reference:

SNM-1097

i June 1, 1984 cylinder is less than 0.90. It is therefore concluded that the inch Schedule stainless steel 304 t cylinder is geometrically safe for 00 2 rod lattices in water.

4.7 CRITICALITY SAFETY OF UPMP SOLVENT EXTRACTION COLUMNS Scrap processing operations in the Uranium Process Management Project employ solvent extraction t (SX) columns to recover purified uranium in the form of uranyl nitrate hexahydrate (UNH) from digest feed '

material with potential metallic impurities or neutron poisons (such.as boron, gadolinium or cadmium). These SX columns have been designed to be geometrically safe '

for homogeneous mixtures of UO2 -and water based upon the following analysis.

4.7.1 Solvent Extraction Column Geometry Figure 4-14 shows an illustration of the most reactive :

SX column design. The vessel consists of a foot t !

high inch Schedule cylindrical pipe connected t i on each end by a large disengaging section. t These sections have an t transition section from.the inch cylindrical pipe to - ']

! a final cylindrical .

tank inches high, 3 i inches in diameter and inches t

' ~ - - ' ~ ' ~ ~

thick. The disengaging sections are not t

~ neutronically isolated by use of neutron absorber panels and rely only on their geometrical dimensions and stainless steel walls for geometric safety. The upper .

disengaging.section has vent holes (not t shown in Figure 4-14) at the bottom of the transition piece which permit complete t drainage in the event of leakage.

4.7.2 Analytical Methods The generic SX Model depicted in Figure 4-14 has been analyzed using the GEKENO Monte Carlo Code and homogeneous UO 2 and water Hansen-Roach Cross Sections.

Figure 4-15 shows the geometry model and materials used in this analysis. The transition piece in t this configuration was modelled with the'GEKENO '

generalized geometry option and the remaining parts were modelled with regular geometry.

i Page 4-62

Reference:

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

June 1, 1984 FIGURE 4-14 GENERIC SOLVENT EXTRACTION MODEL t

- t t

t t

t 3, t t

t t

t i t t

t t

Page 4-63

Reference:

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June 1, 1984 FIGURE 4-15 SOLVENT EXTRACTION UNIT GEOMETRY MODEL t

t t

. ; 't t

t t

+1 t

t t

-t t

t.

t t

t Page 4-64

Reference:

SNM-1097

i June 1, 1984 4

4.7.3 Results j For this. analysis, GEKENO calculations were performed i for the water reflected SX column as shown in Figure 4-15 for homogeneous UO 2 and water mixtures spanning the range of optimum moderation. The results are listed in i Table 4-32. The maximum neutron multiplication factor is and occurs for a WF H2 O of 0.20. t

' Since the operation of the SX columns yields uranium densit'ies of only gm U/1 in the solvent (which t is equivalent to a 00 water mixture with a WF H O 2 2 much greater than 0.50) the Keff result for t 0.50 WP of H 2 O is an upper limit for the normal' case.

The SX columns are thus geometrically safe for homogeneous 00 2 and water mixtures.

I 4.8 CRITICALITY SAFETY OF UPMP THREE AND FIVE GALLON PAILS 5

Three and give gallon pails are used in the Uranium . ;

Process Management facility to handle heterogeneous *

.UO 2 Pellet scrap and homogeneous UO 2 powder and sludges.

These containers are both restricted to a gross weight of 35 Kg with contents which are normally dry (< 0.05 WF of H2O for powder or pellets) or wet (>0.50 WP of water for sludges). As shown in the following analyses, both

. types of pails are geometrically safe.

4.8.1 .

Container Geometries

~~

Table 4-33 contains a description of the geometry of three and five gallon pails. Both are a maximum 11-1/4 inches in diameter with heights of 8-13/16 inches for r

the three gallon pail and 13-1/2 inches for the'five gallon pail. The actual volumes are 13.7 and 21.0 liters respectively although the volumes resulting from the maximum diameters and heights are 14.4 and 22.0

-liters.

4.8.2 Analytical Methods s

A comparison with Tables 4-7 and 4-8 shows that the five gallon pail is. geometrically safe by virtue of its volume (21.0 liters versus a safe geometry value of 24 liters for homogeneous UO 2 and water mixtures) whereas Page 4-65

Reference:

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June 1, 1984 l

TABLE 4-32 SOLVENT EXTRACTION CALCULATION RESULTS GEKENO WF H,0 Keff 2 a 0.10 0.20 t 0.25 t 0.30 t

,0.40 t 0.50 t TABLE 4-33 THREE~& FIVE GALLON PAIL GEOMETRIES Height Diameter Volume Pail Designation (Inches)_

Top Bottom (Liters)

Three Gallon 8-13/16 11-1/4 10-23/32 13.7 Five Gallon 13-1/2 11-1/4 10-23/32 21.0 Page 4-66

Reference:

SNM-1097

June 1, 1984 the three gallon is slightly larger than a safe volume (13.7 liters versus 13.0 for the safe volume for

' heterogeneous 00 2 and water). In addition to this, explicit calculations have been performed with the GEKENO and GEMER Monte Carlo Codes to demonstrate that the pails are geometrically safe for homogeneous (for the five gallon pail) and heterogeneous (for the three gallon pail) UO 2 and water mixtures. The geometry specifications used in these calculations are listed -in Table 4-34. The pails were assumed to be completely.

surrounded by a least 12 inches of water and the contents were modelled as. homogeneous or heterogeneous UO 2 and' water-mixtures as described in Section 4.1.7 and 4.1.9. The three gallon pail GEKENO calculations were performed with the Hansen-Roach Cross Section sets adjusted for heterogeneous effects.

4.8.3- Results Tables 4-35 and 4-36 show the results of the GEKENO '

calculations for the three and five gallon pails. As summarized in Table 4-37, these containers satisfy the

, necessary requirements specified in Sections 4.1.2 and 4.1.5 and are consequently geometrically safe.

, -4.9' CRITICALITY SAFETY OF UPMP FURNACE BOATS Furnace boats used in-the Uranium Process Management

. Project oxidation furnace operations are inch by t inch by inch rectangular pans which can t contain heterogeneous or homogeneous mixtures of UO, and water. Since the boat thickness of inches Is t greater than the safe geometry infinite slab thickness of 3.6 inches for. heterogeneous mixtures, this

-rectangular pans have been analyzed with the reactivity formula and the heterogenous Kinf, M2 and A- Parameters in Table 4-3. Table 4-38 shows these results. Since the maximum Keff is 0.883, the furnace boats are geometrically safe.

L 4.10 CRITICALITY SAFETY OF UPMP ROLL CRUSHER Oxidized scrap exiting the UPMP oxidation-furnace is dumped from a furnace boat into a slab.hopp'er, through a roll. crusher and into a three or five gallon pail.

Figure 4-16 shows a sketch of this configuration. The geometry of the roll crusher is such that it has an Page 4-67

Reference:

SNM-1097

June 1, 1984 TAB,LE 4-34 THREE & FIVE GALLON PAIL GEOMETRY MODELS Radius (cm) Height (cm)

Pail Description Inner Outer

Inner Outer

Three Gallon 14.370 14.42 23.00 23.10 Five Gallon 14.367 14.420 33.814 33.920

Carbon Steel Pails TABLE 4-35

FIVE GALLON PAIL CALCULATION RESULTS (Homogeneous U(5.0)O 2 and Water Mixtures)

GEKENO GEMER WF H,0 Keff 2 a Keff 2 a 0.05 t 0.086- t 0.10 '

O.1158 .

0.15 t 0.1573 t 0.20 t 0.2183 t 0.25 t 0.2711 t 0.30 t 0.3172 t-0.3486 t 0.35 t 0.40 t

! 0.45 t O.50 t Page 4-68

Reference:

SNM-1097

June 1,-1984 TABLE 4-36 THREE GALLON PAIL CALCULATION RESULTS (Heterogeneous U(5.0)O 2 + H20. mixtures)

'W/F GEKENO Volume Ratio Keff t o

~

1.0 t 2.0 t 3.0 t 4.0 t 5.0 t 6.0 ; t

. _ . . . _ _ . - . . ' ~ ~ ' '

TABLE 4-37

SUMMARY

OF CALCULATION RESULTS FOR-THREE & FIVE GALLON CONTAINER Optimum Moderation Normal Case Accident Case Pail Designation Keff 2 a Keff ! o 4

Three Gallon t Five Gallon t Page 4-69

Reference:

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June 1, 1984 l

TABLE 4-38 OXIDATION FURNACE BOAT NEUTRON MULTIPLICATION FACTORS FOR HETEROGENEOUS U(5.0)O 2 LATTICES IN WATER CALCULATED WITH THE REACTIVITY FORMULA W/F Keff*

Volume Ratio 1.0 0.843 i

2.0 '

t 3.0 t 4.0 t Y

5.0 1 6.0 t

12 Inch Water Reflected 1

f Page 4-70

Reference:

~SNM-1097

June 1, 1984 FIGURE 4-16 OXIDATION FURNACE BOAT DUMP STATION t

t t

t

- ; t t

t t

t ,

t

t t

t t

t t-t t

t t

Page 4-71

Reference:

SNM-1097

1 June 1, 1984 active volume of less than 4-7/8 inches by 9 inches by 19 inches (13.7 liters). This is a geometrically :1}i for heterogeneous 002 and water as shown by the I

reactivity formula results in Table 4-39. (Kinf, M2 and A parameters used in this analysis are those in Table 4-3).

4.11 CRITICALITY SAFETY OF PROCESS SUMPS AND FLOOR-BASINS :

Liquid' spills and vessel' overflows in the UPMP fluoride waste, nitrate waste, rad waste, dissolver, solvent extraction and UNH concentration areas are contained in area floor basins,and collected in process sumps for pumping back into waste stream process vessels. These basins and sumps normally do not contain significant volumes of liquids'with-uranium densities in excess of the minimum critical value (> . gm U/1), but they t have been designed to be critically safe for more extreme conditions. These conditions and the applicablg criticality safety analysis are presented in the '

following sections.

4.11.1 Geometries The' floor basin is a slab type area with a maximum depth of inches relative to the surrounding walkway. It l is capable of holding 50% to 100% of the contents of a typical~

vessel and is molded into' the concrete t

. - - - - ~ floor rather than created by curbs above the floor' level. The process sump is a inch Schedule i inch OD and inch thick) inch t

. deep stainless steel pipe imbedded in the concrete floor and covered by a. inch deep, inch diameter t cylindrical grating..

4.11.2 Analytical. Methods The process sump floor basin configuration has been analyzed with the GEKENO Monte Carlo Code for homogeneous mixtures of 00 2 and water. The credible accident conditions assumed in this analysis are:

Optimum moderation of-the U0 2 and water mixture and nominal ( inch thick) water reflection on the top. t Full reflection by water is not assumed for the condition because such a condition is not credible for Page 4-72

Reference:

SNM-1097

9

' June 1, 1984 TABLE 4-39 ROLL CRUSHER CALCULATION RESULTS FOR HETEROGENEOUS U(5.0)O 2 & WATER MIXTURES ANALYZED WITH THE REACTIVITY FORMULA W/F Volume Ratio 'Keff 1.0 0.837

.. 2.0 ; t

, 9 3.0 t 4.0 t 5.0 t 6.0 t

. _.._ ._ 8.0 0.824 10.0 0.802 i

l i

[

l-l t.

f l

l Page 4-73 l -.

Reference:

SNM-1097

June 1, 1984 the large floor basin areas and uranium masses and densities involved.

A fuel mixture with a 0.501 weight fraction of water and with full (12 inch thick) water reflection on top. <

Figure 4-17 shows the GEKENO Geometry Model. The following specific cases were analyzed:

~

(1) Isolated floor basin with optimally moderated fuel and nominal water reflection on top.

(2) Isolated floor basin with a fuel mixture having a 0.50 (and greater) weight fraction of water and with full water reflection on top.

(3) Isolated process sump with optimally moderated fuel and full water reflection on top. - i, (4) The full process sump-floor basin configuration with optimally moderated fuel and nominal water reflection on top.

(5) The full process sump-floor' basin. configuration with a fuel mixture with 0.50 (and greater) weight I fraction of water and full water reflection on top.

- ~4.11.3- Results

.The results of these GEKENO calculations are listed in Table 4-40. These are summarized as follows:

(1) The maximum Keff for the nominally reflected isolated floor' basin with_ optimally moderated UO 2 and water is . It is thus t critically safe for this condition.

(2) The' maximum Keff for the fully reflected isolated floor basin with UO 2 and water mixtures with weight fractions of 0.50 or more is The t isolated floor basin is critically safe for this condition.

(3) The maximum Keff for the fully reflected process sump with optimally moderated 00 2 and water is T

a Page 4-74

Reference:

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June 1, 1984 FIGURE 4-17 GEKENO GEOMETRY MODEL FOR PROCESS SUMP-FLOOR BASIN ANALYSIS t

t t

- ; t t

t t

t Page 4-75

Reference:

SNM-1097

June 1, 1984

-TABLE 4-40 UPMP PROCESS SUMP & FLOOR BASIN CALCULATION RESULTS (Homogeneous U(5.0)O 2 & Water Results)

GEKENO WF H 3 0 Keff 2 o Case A. Isolatbed Floor Basin 0.10 with Nominal Water 0.20 t Reflection on Top 0.25 t 0.30 t 0.40 t 0.50 t Case B. Isolated Floor Basin 0.50 t with Full Water 0.60 t

Reflection on Top 0.70 t Case C. Isolated Process Sump 0.10 with Full Water 0.20 t Reflection on Top 0.25 t 0.30 O.40 .

0.50 t Case D. Process Sump-Floor 0.10 t Basin Configuration 0.20 t with Nominal Water 0.25 t Reflection on Top 0.30 t 0.40 t O.50 t Case E. Process Sump-Floor 0.50 t Basin Configuration 0.60 t with Full Water 0.70 t Reflection on Top l

Page 4-76

Reference:

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June 1, 1984

, the process sump by itself is t geometrically safe.

(4) The maximum Keff for the optimally moderated configuration with nominal water reflection on top is and-occurs for a UO 2-H O t mixture with a weight fraction of water of b.25.

The process sump-floor basin is thus critically safe for this condition.

(5) The maximum Keff for the fully reflected process sump-floor basin configuration with 0.50 or more

. water is and is critically safe t ,

for this condition.

4.12 CRITICALITY SAFETY OF UPMP EQUIPMENT INTERACTION Criticality safety for interaction between individual 4

pieces of equipment in the UPMP facility will be - i demonstrated-for optimum or maximum credible interspersed moderation between units. The methods used and areas of application are as follows:

4.12.1 Monte Carlo Code Applications GEKENO and GEMER Monte Carlo Codes will be used in the analysis of close-packed interacting systems with regular' units or close spacings. Typical analyses will (1) filter area t .

(2) Furnace queue conveyors (with three or five gallon pails)

(3) vessel piping (general) t (4) Discharge furnace conveyors (with three gallon Pails)

(5) Dissolver queue conveyors '(with three gallon pails)

'( 6 ) piping t (7) Rad waste tank piping t (8) Leacher area (9) Oxidation furnace boat dump station 4.12.2 Solid Angle Code Applications- ,

In selected cases, criticality safety of interacting units in the UPMP facility are demonstrated using the Solid Angle Code (SAC). Typical examples are:

4 i

5 J

Page 4-77

Reference:

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4 June 1, 1984 (1) Solvent extraction cell (2) Oxidation furnace (3) Dissolver area For application in the Solid Angle Code, neutron multiplication factors for the unreflected units are computed assuming normal case geometries but optimally moderated fuel (i.e., 00 2 and water) mixtures. These unreflected Keffs are calculated with GEKENO or GEMER Monte Carlo Codes or with the reactivity formula using the Kinf, M2 and A parameters listed in Table 4-41.

Table 4-42 summarizes unreflected Keff values for typical units. Units in solid angle calculations are in addition required to be separated by at least 12 inches edge-to-edge. Closer spacings require analysis by other methods such as Monte Carlo.

4.12.3 Isolation By Distance Equipment'and operations in the UPMP facility are '

considered to be nuclearly isolated if they are' separated from all other accumulations of fissile material by_ distances which are equivalent to the isolation provided by an eight' inch thick slab of water or 12 feet edge-to-edge separation.

4.12.4 Isolation of Process & Dissolver Vessels As noted in Sections 4.5 and 4.6, the use of. neutron absorber panels in process and dissolver .

vessel designs is such that infinite planar arrays of the vessels have neutron multiplication factors less-than 1.0 under optimum moderation and full (top and bottom) reflection. While not nuclearly isolated, such

, vessels in the UPMP facility are considered to have t

negligible interaction (due predominantly to the neutron absorber panels) and are not usually t considered in interaction analyses. Monte Carlo analyses referenced in 4.12.1(3) and 4.12.1(6) also

, demonstrate generically that . vessels and their t l t associated piping are critically safe. j i

Page'4-78

Reference:

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, - - , , . - ,.-,.,-e

June 1, 1984 TABLE 4-41 REACTIVITY FORMULA PARAMETERS FOR UNREFLECTED UNITS

FOR USE IN THE SAC SOLID ANGLE CODE M2 x Material Type Kinf (cm ) 2 (cm)

Homogeneous 1.442 28.0 2.1 Heterogeneous 1.540 29.7 3.2

U(5.0)O 2 and H 2 O Mixtures 3,

4 TABLE 4-42 UNREFLECTED KEFFS FOR SOLID ANGLE CALCULATIONS UO 2 & Water Equipment Mixture Form Keff Source Three Gallon Pails Heterogeneous 0.715 GEKENO Five Gallon Pails Homogeneous .0.767 GEKENO Inch Schedule t Cylindrical Tanks Homogeneous GEKENO t Leachers Heterogeneous GEMER t Solvent Extraction Homogeneous GEKENO t Columns calculation I

Furnace Boats Heterogeneous 0.618 Reactivity Formula Roll Crusher Heterogeneous 0.678 Reactivity Formula I

I i i l

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..- - , . , , . _ , . - .. , , , . . _ . . . , , . . , - - - , , , .-1

June 1, 1984

~

4.12.5 Criticality Safety of UPMP for Process Piping Interaction In the UPMP facility, process piping has been designed to be critically safe by implementation of the following generic guidelines.

4.12.5.1 Interaction of process piping in excess of 2 inches in diameter and which normally contains levels of uranium in excess of 1,000 ppm must be specifically analyzed using Monte Carlo reactivity formula or Solid Angle Code methods. Subject to the additional guidelines below,

^

piping-involving diameters of 2 inches or less, or uranium densities of less than 1,000 ppm do.not require a specific interaction analysis.

4.12.5.2 Minimum spacing of process piping from other piping and equipment containing uranium is shown in Table 4-43.

4.12.5.3 . Pipe runs, where possible, are to be routed in planar slabs with slab thicknesses of no more than 4 inches.

At least a 3. foot surface-to-surface spacing is required when two or more of such pipe runs are stacked on top of each other.

4.12.5.4 Piping to equipment is to be designed so that individual lines and connections are separated as far as practicable from other lines and connectors.

~

4.12.5.5 overflow lines, vents, etc., which do not normally contain uranium bearing materials are not considered in area interaction analyses. These are to be spaced as far away as practicable from uranium bearing equipment and process piping.

Page 4-80

Reference:

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June 1, 1984 TABLE 4-43 SPACING GUIDELINES FOR UPMP PROCESS PIPING Normal Required Equipment Minimum Spacing, From To (Inches)

>2" Pipe Other Piping, Equipment 12" or vessel 3

>3" Piping Other Piping, Equipment 18" or vessel

>4" Piping Other Piping, Equipment 24" or vessel

- Angled Takeoffs Other Piping, Equipment 15" Process Piping or Vessel Page 4-81

Reference:

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June 1, 1984 CHAPTER 5.0 RADIOLOGICAL SAFETY CONSIDERATIONS 5.1 l BUILDING VENTILATION SYSTEMS l l

The UPMP heating, ventilating, and air conditioning systems are provided to maintain the temperature, humidity, air cleanliness, ventilation rate, and contamination control environments required within the facility. Temperature and humidity are controlled as required for the processes performed in the areas served, and for the efficient performance of operating.

personnel. The release of radiological and non-radiological particulates, aerosols, fumes, and vapors is controlled to as low a level as practical by

' filtration through low, medium, and high efficiency filters and by scrubbing the air with deionized water.

The UPMP controlled areas are maintained at a negative '

pressure with respect to atmosphere and adjacent areas

.(with the exception of the GECO vaporization room which shares a common wall with no openings between it and UPMP). Because the vaporization area is considered as having the higher potential for contamination, it is maintained at a higher negative pressure. Automatic room static-pressure sensors continuously modulate control dampers-to maintain the required pressure differentials between adjacent areas and the environment.

5.1.1 Recirculation Systems Separate recirculation systems are used to' segregate the main process and solvent extraction areas. All

!; recirculated air passes through high efficiency particulate air (HEPA) filters and is scrubbed prior to

! being returned to the work areas. . Clean recirculated supply air. enters those sections of the room with the least potential for contamination. Returns are located so that the air flow is directed away from areas normally occupied by personnel.

l l

i I

i l

l l

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Reference:

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

~

June 1, 1984 4

5.1.2 Equipment Enclosures Special enclosures are provided to supplement.the primary containment offered by the process equipment.

Enclosures are provided for each piece of process equipment where a potential for airborne contamination exists. Each enclosure has been designed to minimize the potential for direct contact of radioactive material by the, operator and to maintain exposures ALARA.

An automatic air bypass damper is provided on each exhaust duct to maintain a constant negative pressure 4

within the enclosure. By properly adjusting this damper, the required air flow through routine access openings (such as those required for material transfers or cleanouts) is provided. Under normal, and most anticipated abnormal process conditions, respiratory protective equipment will not be required.

5.1.3 Absolute Filter Systems

. Exhausted air from enclosures is filtered by a HEPA filter provided for each dry containment system. .

Exhausted air is further filtered by a secondary bank of HEPAs prior to discharge to the environment.

The primary and secondary absolute filter systems consist of ~40% efficient, fiberglass prefilters followed by the HEPA filters. The HEPA filters are 99.974 efficient for 0.3 micron and greater dust particle sizes.

The filter housings are' designed such that the fiberglass prefilter is totally sealed and cannot be bypassed, thereby assuring that all entrained particulate material must pass through the prefilter before-reaching the HEPA filters. The HEPA filters are enclosed on all four sides with spacer material to

, assure that no buildep of material.can occur in the perimeter of the housing.

All filter housings are equipped with a magnahelic gauge to measure the pressure differential (A P) across the filters. On the secondary filter. banks, a pneumatically ~

controlled damper is used to automatically adjust the airflow through the filters. As the HEPA filters load Page 5-2

Reference:

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June 1, 1984

with particulate material, the damper automatically opens to compensate for the increased A P. When~a differential pressure is observed to have reached four inches of water, the effectiveness of the system is-investigated. A filter is replaced following avidence of the inability of the exhaust system to perform its function properly or maintain minimum inward airflow.

The filters are never operated at a P values which exceed the manufacturer's ratings.

5.1.4 Ductwork

' Ductwork for the exhaust and recirculation air main '

trunklines have continuous welded longitudinal seams and companion-angles with gaskets at all joints.

Horizontal ducting between process enclosures and primary filters has been kept to a ' minimum, at. there utilized, the duct has been equipped with clea . vat openings for inspection and removal of uranium . ; ;

accumulation in the unlikely event it occurred. '

The building ventilation major equipment.and specifications are listed in. Table 5-1.

5.2 EQUIPMENT OPERATION .

5.2.1 Facility Exhaust System This main exhaust system draws air from four different areas, treats the air to meet emission requirements and monitors the air as-it is exhausted. The four air collection areas are:

o Dry hoods (hoods which contain a powder) o Wet hoods (hoods which contain a slurry or liquid) o Rooms which require a high make-up ratio (solvent extraction and dissolution rooms) o Main process area 5.2.1.1 Dry Exhaust System Each of the dry hoods is provided with an exhaust air duct connection and a bypass air spoiler damper.

Exhaust air flow is through a primary HEPA filter on the exhaust piping from each individual hood thus minimizing balancing problems which occur when primary HEPA filters are shared between hoods. The primary HEPA filters are connected in parallel and ducted to the downstream. side

~

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Reference:

SNM-1097

--w,--- - - - - - -

,, n- ,e .- ,= - , , - - - - , . - - - - - ,,--,,n, , _ , - , _ , _ . . _ . , - - , - . - - . , - . - --, - -

June 1, 1984 TABLE 5-1 BUILDING VENTILATION SYSTEMS MAJOR EQUIPMENT LIST BLOWERS Process Area-Recirculation Blower (B-950)

+

45,500 cfm Process Area Exhaust Blower (B-965,966) 16,000 cfm ea Solvent Extraction Recirculation Blower (B-940) 20,000 cfm

' Process Lab Blower (B-970) 3,200 cfm Control' Room Supply Blowers (B-975, 976) 5,750 cfm ea Equipment Room Supply Fans (B-985,986,987,988) 10,000 cfm ea UPMP Corridor Supply Fan (B-2016X) 12,000 cfm FILTERS

Process Area Recirculation Filters (F-950) 32 Unit Process / Scrubber-Exhaust Filters (F-965, 966) 16 Unit Solvent Extraction & Dissolver Filters (F-940) 20 Unit-Process Lab Filters (F-970) 2 Unit Control Room Filters (F-975) 6 Unit l

_ _ SCRUBBERS _

Process Area Recirculation Scrubber (S-950) 200 gpm Process Area Exhaust Scrubber (S-965) 120 gpm Solvent Extraction and Dissolver Scrubber (S-940) 140 gpm HEATING / COOLING COILS-Process Area Exhaust Scrubber Coils (HV-968) 1 Process Area Recirculation Coils (HV-951,952)

Solvent Extraction and Dissolver Room Coils (HV-941, 942)

Process Lab Coils (HV-97.1, 972) t Page 5-4

Reference:

SNM-1097

O N

- 'Jun's 1, 1984 s:

of the scrubber reheat coil. From here the air travels through one of the redundant sec6ndary HEPA filters and 3 ita corresponding exhaust fan.

' Pressure in'each hc6d relative to the room is sensed by a pressure differential transmitter and a pneumatic signal is transmitted to the pressure differential controller. The spoiler damper is modulated as required to maintain the hood-at a negative pressure of 0.2 to 0.75 inches of water with respect to the room. If pressure in the hood increases and becomes greater than

-0.2 inches of water, the spoiler damper will close to insure that there is a minimum of 80 LFPM at all points with an average of at least 100 LFPM air velocity into '

any routine hood' opening including open doors, slides, etc. A pressure differential indicator is provided- at each hoed. -

5.2.1.2 Wet Exhaust System ;

The. wet e.'hau'st t system handles the furnace scrubber, dissolver, le'acher, process filterrehelosure, filter hood, and the tanks. An NO x absorbe'r and blower are provided in the facility for that e.quipment emitting gases with a high NO x content. A' manual damper is provided for small hoods. The wet hoods are connected in parallel and ducted to the exhaust scrubber. The exhaust scrubber sprays 12.0 gpm of recirculated, acidified, de-ionized water'into the airstream via a 4

' foot thick packing section primarily to remov'e'non-radiological particulates'and fumes. The air leaving the scrubber passes through a heating coil to decrease the relat-ive humidity of the air prior to passing through the second,ary filter, thus keeping the secondary filter dry. a 5.2.1.3 High Make-up Ratio Rooms ,

Thesolventextractionanddisdolutionroomsare provided with a air turnover rate of at least eight room air changes per hour. The dissolver hoods are

' located in the. dissolution' room on the mezzanine and thus the air flow to the hoods can provide room exhaust.

The exhaust air from these rooms Ls adjusted by

automatic room prescure control dampers.

__ 't d *

'M Page 5-5 Reference; SNM-iO97 <

June 1, 1984 The air is ducted to the scrubber in a similar manner to that being used for the wet hoods.

5.2.1.4 Main Process Area The main process area is provided with an air turnover rate of at least eight room air changes per hour.

Discrete confinement zones are provided using localized supply and return openings to minimize.the consequences of an dirborne release. Main overhead supply ducts provide clean, recirculated air above the center walkways. High and low return openings are provided at various locations behind the process tanks along the '

perimeter walls to minimize airborne releases and direct the air flow away from the operators. In addition, under normal. operating conditions, process enclosure access . openings will be closed so that .the facility exhaust system will be exhausting a significant portion of room air.via the bypass dampers on each enclosure.- (

5.2.2 Solvent Extraction & Dissolution Room Make-up &

Recirculation System This system maintains the req'uired air temperature and provides make-up and air purification for the solvent i extraction and dissolution rooms. Air make-up is 3 provided from the main process area and is equal to the air being exhausted from the room. Individual pressure control devices and air operated dampers are provided to maintain the proper negative pressure by regulating the exhaust air rate from each room.

The recirculated air is passed through a HEPA filter, cooled, scrubbed and heated, if necessary, and then sent

through.each room. The system is provided with control devices and dampers to maintain a constant flow through the system. A pressure differential indicator and alarm switch are provided on the secondary HEPA filter.

5.2.3 Process Area Recirculation System The process area recirculation system consists of a 32 unit HEPA filter, cooling coil section scrubber system, reheat coil, recirculation fan, supply air ductwork and automatic controls.

i l

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Reference:

SNM-1097

_ , _ . . - _ _ _ _ , - _ . _ , _ _.-_._m.,__ _ , - -- - . - - - . , v-- , -- w.----. ---- . - - - - - - - - - - - - - - -

y

- 8

, v June,1, 1984 The operation of the recirculation air fan is started

, from the control room. The recirculation fan flow rate is sized tx) provide a minimum of eight air changes per hour. This flow, along with the make-up air, is capable of cooling the process area. The recirculated air

_ passes through a filter plenum and into the HEPA filter units The air is passed through a cooling coil scrubber and reheat section where the air is simultaneously, b washed with de-ionized water, cooled 'and dehumidified to obtain[the specified room dew point t'emperature. '

The recirculated air is next reheated and returned to the process area. The system is provided with control devices and dampers to maintain a constant flow rate through the main recirculation duct. A pressure differential indicator and alarmcswitch are provided for the secondary HEPA filter system..

Cooling, heating and humidity control are provided by. ; ' k' the sprayed coil dehumidifier and the reheat coil.- A '

temperature transmitter located in the process area

. transmits a pncumatic signal /to a receiver controller which in turn modulates a three way chilled water valve and a two way steam control! valve to maintain desired room temperature. Recirculated deionized water is

, continuously sprayed over'the downstream side of the chilled water coil to scrub the air stream. The water level in the recirculation sump ic maintained by a. 1 controlled volume of DI water, i; is volume of make-up_

water is determined by the c74)cr tion rate and an 11 liter / minute-purge to the se,1;O.ust scrubber. This will ensure scrubber efficiency c.t3 prevent contaminants from collecting in the sump.

Anionizationtypesmokedetebtorislocatedinthe return air duct upstream of the HEPA filter plenum. On detection of smoke,.a remote alarm is energized in the control room and kthe system is shut down.

5.2.4 Process Area Make-up Air System A separate outside air system is provided to supply make-up air _to-the process' area. The system provides 100% outside air and incorporates a roughing filter at the roof intake, cooling coils, steam heating coils, f, t

+

Page 5-7

Reference:

SNM-1097 ,

June 1,1984 ductwork and controls. The system utilizes existing condensing units. A room pressure sensor is utilized to modulate the control damper to maintain the main process area at the proper negative pressure.

5.2.5 ASMU Operating Area (Non-controlled)

The aqueous and solvent make-up (ASMU) operating area HVAC system will provide fresh air ventilation for the summer'and heating for the winter. The area will be kept slightly positive or at atmospheric pressure since it is a noncontaminated area located adjacent to a contaminated area.

5.2.6 ASMU Chemical Mix Area (Non-Controlled)

Heating and ventilation is provided in t'he bulk storage area. Continuous air make-up is provided to prevent fume accumulation. The unit is supplied with a heater ;

to eliminate pipe freezing concerns.

5.2.7 Control Rmam and Office (Non-Controlled) l l

The control room / office. area / computer room system consists of a return and make-up air plenum, filter l

section, charcoal filter, fan and cooling coil section, l heating coil section, ductwork and automatic controls.

- The supply fan is' energized manually and operates continuously. A redundant fan is supplied to ensure that a reliable system is provided. A manual make-up

air damper is set to allow 240 cfm of.make-up air to mix with 5,600 cfm of HEPA-filtered return air. The 240 cfm

' make-up provides the positive pressure required for these non-contaminated areas. The air is drawn through a steam heating coil, a chilled water coil and sent to the different areas. A carbon filter is provided to remove unwanted' fumes from the atmosphere.

5.2.8 HEPA Filter Banks I

Flow through the recirculation and main exhaust HEPA

' filter banks will be controlled by a damper controller p in order to maintain the required negative static y pressure at each fan suction. In the event that the lj filter bank plugs to the point where the required

?

Page 5-8

Reference:

SNM-1097

.i _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _

t

June 1, 1984 l

r negative pressure cannot be maintained, a controller for the recirculating air units will modulate the fresh air dampers towards the closed position. When a HEPA filter bank differential pressure reaches 4" of water, a visual / audible and printed' alarm will be indexed in the UPMP control room and the effectiveness of the system will be investigated.

4 5.3 PERSONNEL EXPOSURE CONTROLS 4

e The UPMP facility will be operated according to the radiological control plan which' exists for the current I

fuel manufacturing building. This includes exposure controls, personnel monitoring techniques, bioassay i.

programs, area posting and radiation surveys.

5.3.1 Measurement of Air concentrations

- , Work area air concentrations in the UPMP facility will ;

4 be continuously sampled with the addition of 47

-stationary air samplers (SAS). Each sampling unit has a y flow meter with-flow rate control and a filter holder containing a round glass fiber filter which, per existing radiation protection procedures, is changed every shift, or' sooner if conditions warrant.

These filter samples are analyzed each shift for alpha

. activity within approximately four ho'urs following removal from the filter holder. This analysis is performed by placing the filters into individual planchets and loading them into low background alpha counting systems. The total counting time for all filters is approximately two hours.

Individual airborne exposures will be assigned to personnel working in the facility based upon sample results and the amount of. time spent in each work area.

The SAS samplers are augmented by four additional continuous air monitors which continuously monitor air concentrations and provide early warning of containment failure or releases for each area. If a problem does exist, employees in the area will either be' evacuated or required to wear respiratory protection until the problem is resolved. Additional air concentrations.are _

determined using high volume, portable units as necesary A

i-Page 5-9

Reference:

SNM-1097

June-1, 1984

.to determine the temporal nature of air concentrations in accordance with WMD procedures.

5.3.2 Contamination controls Process areas of the UPMP facility are designated as controlled areas similar to the controlled process areas of-the existing fuel manufacturing building. UPMP controlled areas include the main process tank ' area, oxidative furnace area, dissolver/ leacher areas, solvent extraction. cell-and the maintenance shop.

Radiation workers will access the controlled areas through existing change rooms, where they will don

standard controlled area protective clothing (i.e.

coveralls, head covering, shoe covers, rubber gloves).

, Persons exiting the controlled arca will monitor for contamination following removal of protective clothing in the change rooms. - ;

The UPMP processes have been designed to offer complete containment to the' work areas, thus minimizing the potential for surface and airborne contamination.

Liquid effluent streams-are. piped directly into closed process vessels which are vented into the main scrubber exhaust system. Solid waste streams are transferred ;

into and out of'the process areas in closed three and five gallon pails and are not opened outside of approved hoods. Specially designed enclosures are provided for pail loading and unloading.

The existing fuel manufacturing building contamination control plan and action guides will be used for the UPMP facility. If contamination in excess of the guideline limits occurs, the necessary decontamination action is taken per existing procedures, based upon knowledge of the particular circumstances and the behavior of the material' involved.

5.3.3 Criticality Detection & Evacuation Alarm System The UPMP facility is adequately covered using the existing criticality warning system and with the addition ~of one detector in the_ main process area. The alarm signal will be of sufficient volume in all areas to be evacuated with the addition of 21 horns to the

~

i Page 5-10

Reference:

SNM-1097

J June 1, 1984 existing system. The evacuation alarm sound will be consistent with that of the existing fuel manufacturing building. Individuals evacuating the UPMP facility will do so according to evacuation procedures outlined in the existing site emergency plan.

5.3.4 Operating Instructions to workers The' operation of the UPMP facility will be conducted according to Process Requirements and Operating Documents (PRODS) prepared by process engineers with inputs from nuclear safety engineering personnel. These

. ' documents provide on-the-floor instructions to operations personnel and contain criticality and radiological safety provisions. Each equipment operator is provided adequate training to follow these operating

-documents.

3 Page 5-11

Reference:

SNM-1097

June 1, 1984

CHAPTER 6.0 -i ENVIRONMENTAL CONSIDERATIONS l l

6.1 OVERVIEW The UPMP facility will have a positive environmental impact with respect to the facility's currently approved

. operations and with respect to the entire fuel manufacturing cycle. In this regard, the UPMP license amendment should receive favorable-treatment under the provisions of 10 CFR 51.5(d)(4). The activities are included in the current facility license and the project represents only a rearrangement of the current process coupled with modifications and inclusion of new-equipment and processes for which safety has not been demonstrated under the current license. In-support of

.these claims, the following environmental benefits / improvements are demonstrated: . ;

(1).The quantity of uranium in radioactive liquid F (fluoride, nitrate, rad waste) exiting.tne normal-confines of the manufacturing operations and entering the open waste treatment systems will be

, reduced by.approximately %. This eliminates t any major potential for its contribution to a

, radiological impact on the environment.

~ (2) The quantity of calcium fluoride sludge produced and

~ ~~

precipitated in the final process lagoons from rad waste will decrease by approximately %. This t eliminates a significant quantity of radioactive materials generated during fuel production and their attendant environmental impact potential including considerations for site decommissioning.

l (3) The uranium concentration of the remaining calcium

fluoride sludge will be signicantly reduced I resulting in a non-hazardous waste which can be considered for disposal under Options 1 and 2 of SECY 81-576 or chemical uses in other non-nuclear operations. This eliminates the need for disposing

of current generation material at a low level radioactive waste burial site.

i-Page 6-1

(.

Reference:

SNM-1097 l.. . - . - . . . _ - . - , . - - _

June 1, 1984

-(4) These process modifications make the WMD fuel manufacturing operations more self-supporting at a single location thereby eliminating over-the-road shipment of scrap for reprocessing and return shipment of UNH solutions. This eliminates the potential of a release to the environment during an accident on the highway.

(5) The new process chemistry will eliminate the use of hydrogen peroxide'in the current UPS. This eliminates a potentially hazardous industrial chemical operation.

(6) The UPMP process design emphasizes resource a conservation. Solutions and chemicals associated with the process are treated so as to permit recycling within the operations. This results in a reduction in the quantity of waste chemicals and the need for their disposal. .

(7) UPMP provides the first step in developing the.

< capability to recover uranium from past plant effluent residues. This is an essential part of the GE Closure & Decommissioning Plan for _

the Wilmington Manufacturing Department facility.

(8)~Notwithstanding the fact that UPMP will add a single atmospheric discharge point, the overall discharge

" -~~-~

to the atmosphere from the' fuel manufacturing cycle is expected to be _ reduced. At the WMD site, the atmospheric discharge will be insignificantly increased, however, the atmospheric discharges from reprocessing WMD scrap will be eliminated at the current reprocessing facility. The net result should be a favorable impact on the fuel manufacturing cycle environment.

(9) UPMP will result in a slight increase in the nitrate-based chemical processing at WMD.

Offsetting this is the fact that nitrate-based chemical processing will be eliminated at another facility. .This slight increase at WMD poses no increased environmental risk because of the biological disposal option in use in this area.

4 1

Page 6-2 Referencer SNM-1097

=_. . . - - . .. - - - - , . . . _- -

June 1, 1984 The following segments of UPMP will contribute to the overall improvement in WMD environmental quality:

Fluoride waste'tre'atment Rad waste treatment Nitrate waste treatment Scrap processing UNH conversion The effect of each of these segments on releases to the environment is discussed in detail in the following section.

'6.2 EFFLUENT QUALITY 6'2.1

. Treated Process Liquid Effluents The existing process liquids are segregated by. chemical characteristics prior to the treatment processes thus - i,

-providing the opportunity to achieve optimum removal efficiencies. A flow diagram portraying this segregration and the overall process effluent flow path is provided in Figure 6-1. Figure 6-2 indicates how the Uranium Process Management Project segments fit into the existing treatment concepts.

6.2.1.1 Fluoride Waste Treatment The UPMP units (which have already been t tested) are designed to reduce the uranium chemical concentration to part per million prior to the t existing waste treatment steps. This represents a significant reduction in uranium concentration from the present range of. parts per million at this point t in the waste treatment. process. As a result, the residual concentration of uranium in the lagoon systems will be reduced by a factor of and the calcium t fluoride solids will be less than parts per million t on an as-produced basis orless than pCi/gm at 4% t enrichment. This' low activity concentration will allow consideration of disposal options 1 or 2 as detailed in SECY 81-576. This will also allow consideration of other beneficial uses of the material or, in any case, 4 disposal at a burial facility other than one licensed to receive low level radioactive waste. It should be pointed out that the calcium fluoride solids are not i

1 Page 6-3

Reference:

- SNM-1097

June 1, 1984 1

FIGURE 6-1 l PROCESS EFFLUENT FLOW PATH I PEATING m WASTE AIRCRAFT ENGINE PRECIPITATION AND EQUIPMENT ETCH ' TREATMENT MANUFACTURING WASTE BUILDING SPENT OFFSITE NITRIC DISPOS AL -

SPENT ETCH 1 r OFFSITE FUEL DISPOSAL COMPONENTS OR USE OFF MANUFACTURING U

N SITE BUILDlHG )

USAGE SPENT SODIUM '

HYDROXlDE FINAL DISCHARGE n NITR ggg7gATE . SAMPLE POINT %

A' PRECIPITATION gk ANUFACTURING BUILDING TREATMENT j JLUDGEl FOR NITRATE LAGOONS TEWORK l

. FLOORIDE

" WASTE l z g

AMMONIA p SLUDGE 4 pl r BEPARATIOPI RECOV 5 4 SLUDGE lFOR FLUORIDE 9[<

REWORK 1P E

m RECOVERED LAGOONS j AMMONIA fj RAD W ASTE AERATION LAGOONS e 7

N ADJUST FINAL PROCESS LAGODNS FINAL DISCHARGE SAMPLE POINT DAM vd l

TO RIVE l

l i

Page 6-4

Reference:

SNM-1097

June 1, 1984 FIGURE 6-2

, WILMINGTON SITE PROCESS EFFLUENTS & TREATMENT PEATING ,

AIRCRAFT ENGINE A TE '

AND EQUIPMENT ETCH

'^

MANUFACTURING WAST [

BUILDING SLUDGE SPENT NITRIC OFFSITE Dispos AL I I

SPENT ETCH

OFFSITE i

' FUEL DISPOSAL COMPONENTS OR USE MANUFACTURING N OFF BUILDING O

) SITE .

SPENT SODIUM USAGE .

HYDROXIDE REACTION FINAL DISCHARGE

~

L4 SOLIDS REMOVAL SAMPLE PO!NT FUEL Y ILD NG N /

NITRATE LAGOONS h E

+S OLID S (b U FLUORIDE TO BUR!AL i g-EACTIO RECYCLE. vm w3sye i g

AND LUORIDE AMMONIA _p SOLIDS REATMEN RECOV Ol 5

EMOVA ME s I s

/\b O/ tu' SCRAP ' o 1F FLUORIDE

PROCESS 2 RECOVERED LAGOONS IE W '

AMMONIA J

f, $

RAD WASTE 7 AERATION LAGOONS e

\ ADJtJST FINAL PROCESS LAGODNS FINAL DISCHARGE SAMPLE POINT -

KEY: "%*E UPMP PROCESS '- '

DAM STEPS vd TO RIVE Page 6-5

Reference:

SNM-1097 :

J

June 1, 1984 expected to be a hazardous waste under the criteria of the Resource Conservation and Recovery Act regulations.

The liquid remaining after treatment has a uranium concentration of less than one part per million, equivalent to present day values.

The residual fluoride concentration in the treated liquid will not change from present day values.

6.2.1.2 Rad Waste Treatment The installation of the chemical treatment capability for uranium precipitation and multi-stage solids removal 8

will result in a lower residual uranium concentration in the treated rad waste liquid entering the final process lagoons. The concentration is projected to be less than parts per million, substantially less than the t present range of. parts per million. The sludge t produced in the treatment will be recycled to scrap -

i, processing thereby reducing the quantity of sludge reaching the final lagoon system and the quantity of urenlum precipitated in the final process lagoons. This chemical treatment will also reduce the amount of fluorides discharged to the final process lagoons.

6.2.1.3 Nitrate Waste Treatment l 4 The installation of the additional reaction vessels and solids removal equipment is expected to reduce the concentration of uranium discharged to the nitrate lagoons to parts million, less than 5% of t present day values. In addition, no sludge will be expected to be diccharged to the nitrate lagoons due to the increased effectiveness of the chemical treatment and solids removal processes.

The overall volume generation rate of nitrate waste is

. not expected to change significantly. Any changes in composition are also expected to be minor. In any event, these minor changes will be well within the assimilative capacity of the receiving treatment facility.

Page 6-6

Reference:

SNM-1097

7 June'1, 1984 6 ,2" .1. 4 Scrap Processing

_/ The scrap processing operation will not have an impact

,/ on the treated process effluent discharged from the

/ ' site. The scrap processing portion of UPMP can be

/

/: separated into three separate major chemical operations:

oxidation, dissolution and extraction. The gaseous

", effluents from the oxidation operation are treated by wet scrubbers and condensers with subsequent release through the single UPMP exhaust stack. The gaseous effluents from the dissolution process contains oxides

.of nitrogen at the vessels. These are passed through demister pads, condensers and finally through a NO;.

absorber which reduces the nitrogen oxide levels prior to release through the exhaust stack. The discard liquid streams from both these operations are routed to

.the nitrate waste system for further treatment. The insoluble fraction from the leaching operation will be packed for offsite disposal. These solid wastes are . ;

discussed in Section 6.4.2.

o The product, purified uranyl nitrate, is transferred to the product storage tanks from which it is processed through a dedicated conversion operation which produces ceramic grade U0 2 powder.

o The solvent stream is pumped from the strip column to the solvent treatment system wash where an ammonium j~ ' ~

carbonate solution is used to strip the solvent of the organic decomposition products. The clean organic stream is cecycled to the extraction columns while the waste products (~ liters / week) are t sent to the on site incinerator for disposal.

o The extraction column aqueous waste (AW) liquor contains all of the metal impurities which were extracted from the feed stream plus an insignificant amount of solvent / organic carryover. The AW stream is processed through the secondary nitrate waste treatment operation in which it is lime treated and the resultant precipitate, which includes the heavy metal ions as well as the trace residual organics, is filtered and packed for subsequent off site disposal.

Page 6-7 Reference SNM-1097

l

-June 1, 1984 6.3 AIRBORNE EMISSIONS

'All process emissions from the Uranium Process Management Project will be discharged through one new exhaust stack. The three effluent constituents of potential concern (uranium activity, fluoride and nitrogen oxide concentrations) will be continuously sampled or monitored as is described in .the current facilit,y license information.

The uranium activity and fluoride concentration will be determined based on a continuous sampling system similar to those already serving the existing chemical processing area exhausts.- The nitrogen oxide -

concentration will be continuously monitored at the

. discharge of the NOx absorber.by an on line nitrous oxide (NO) analysis. The monitoring unit i.s catalyst based and measures the total NOx content of the gas stream by converting the NO 2 to NO. The NO 2 -

contribution can be determined by a difference technique' by periodically bypassing the catalyst.

It is estimated that the uranium activity in the exhaust stream will be about 0.05 x 10-12 uCi/cc. This activity is about one third the 1983 average I concentration of 0.14 x 10-12 uCi/cc for all stacks on I the site and would be expected to cause less than a two percent increase in total activity released from the site. This value is conservative since it includes no -[

consideration for the reduction in the current exhausts resulting from the discontinued operation of UPS.

This processing activity replaces the same type of processing activity at another reprocessing facility with newer, modern technology. The net effect is that emissions from the fuel manufacturing cycle ~will be reduced slightly.

The fluoride releases are anticipated to be less than 100 grams per week. This quantity would equate to a fluoride concentration of 2.5 x 10-3 ugms/M 3 at the site boundry using conservative meterological assumptions.

This concentration is well below the generalized threshold of~1 ugm/M 3 for effects on sensitive species found in the National Academy of Sciences Biological Effects of Atmospheric Pollutants - Fluorides.

Page 6-8

Reference:

SNM-1097

June 1, 1984 The-nitrogen oxide releases are anticipated to be 0.7 pounds per' hour. This will result in an increase in the mean annual average ambient air NO 2 concentrations as a result of the Uranium Process Management Project of less than 0.003 ppm. This value, 6% of the federal air quality standards, is obtained by averaging the projected maximum annual NO 2 concentrations in each of

' the sixteen sectors surrounding the site. While representing a slight increase for WMD operations , it actually results in a decrease in discharge as a ,

function of nuclear fuel produced.

The methodology for determining this-concentration is based on the US Nuclear Regulatory Commission Regulatory Guide 1'.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooler Reactors."

The meteorological data used for the calculations was. ;

obtained from the New Hanover County Airport for the '

. period 1978-1982 and is representative of the GE-WMD site. .The data tabulation was performed by the National

. . Oceanic and Atmospheric Administration, National Climatic. Center, Asheville, N.C. These are the same i=

data as used in site evaluations of the current l license.

I 6.4 SOLID WASTE 6.4.1 Sludges from Fluoride, Rad Waste & Nitrate Wastes i

The significant benefit realized from the Uranium Process Management Project has been detailed in Sections 6.2.1.1 (Fluorides), 6.2.1.2 (Rad Waste), and 6.2.1.3 (Nitrates).

In summary:

o The accumulation in the final process lagoons of sludges from the radwaste stream will be eliminated, o It is anticipated that the ongoing accumulation of uranium bearing sludges in the nitrate lagoons will be reduced to less than 4 of the present t quantities.

J t

Page 6-9

Reference:

SNM-1097 f

- - . - - , - , ---,.,,,,.-,n v. __,-,,.,,n.-. _ _ n,.,._ -, _-na,,,.n .,,,,,,,,v-,-- - - . , , _ , , , , - , - .v,--

June 1, 1984 o The disposal of calcium fluoride sludge disposal at other than a NRC licensed burial facility, as well as alternate uses of the material, can now be considered.

- S.4.2 Other Solid Wastes & Sludges The processing of uranium bearing materials in the form of sludges, powder, ceramics and ash at the UPMP facility results in the generation of moist filter cake and solid waste products.

Prior to the implementation of the UPMP facility, all ash and selected discrepant material, much in the form a of sludges, were transported via intrastate trucklines l to the uranium processing facility in Tennessee. The residues from processing this discrepant material were disposed of from that facility. With the implementation of the UPMP facility, the potential for these discards ;

is shif ted to WMD but does not represent any increased '

discard for the fuel manufacturing cycle. This discarded material is composed of three different types:

(1) The insoluble fraction from the leaching operation.

The major contributor is the incinerator ash l resulting from the burning of HEPA_ filters and other combustible wastes generated on the site.

The resultant material should not exceed 50 metric tons per year of moist, liquid-free filter cake

~

. containing less than 0.1% uranium. l (2) The sludge precipitated from the treatment of the solvent extraction aqueous waste stream in the secondary nitrate waste treatment operation. This treatment process output should not exceed 80 metric tons per year of moist,. liquid-free sludge containing less than 0.05% uranium.

(3) The contaminated, non-combustible scrap and debris generated during the day to day, normal operation of the facility, similar in characteristics to the type of material generated in the present fuel fabrication operations.

i l

l I

L Page 6-10

Reference:

SNM-1697 l

9 s .~ > , -<a,w----,,---m -,.e-rn.,. v..e,, e - - - --- - - - - - - -- - - - - - , - m m--n,---- . - -- - - - - , ..-w--,--,--,,,,--w--- er,m-. v -s-g,v -w

June 1, 1984 Because of the improved technology being implemented. in

, the project and the efficiency of operations, these quantities will be less than the quantities produced under the current arrangement.

6.5 REDUCED TRANSPORTATION RISK The provision for on site capability to recover uranium from process scrap and sludges will eliminate the present transportation associated risks from the shipment of uranium bearing sludges and the return shipments of uranyl nitrate. '

6.6 CLOSURE & DECOMMISSIONING The Closure & Decommissioning Plan for the WMD facility is predicated upon complete removal of all sludges and other contaminated materials from the site prior to the closure and decommissioning. The Uranium Process

Management Project provides a key element for '.

accomplishing this plan.

The disposal of sludges resulting from the treatment of

the fluoride waste stream and the treatment of liquids

in the final process lagoons will be facilitated with

this project because it will no longer be necessary- to send these. materials to a low level radioactive waste

' facility for. burial. The Uranium Process Management

. Project also provides one already operational, key element of technology required in the event that the facility is decommissioned.

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Reference:

SNM-1097

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ML20138G235

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