Text

American Centrifuge Plant - Supplemental Proposed Changes for American Centrifuge Operating, Llc'S License Amendment Request (Enterprise Project Identification Number: L-2020-LLA-0085)
	ML21085A524
Person / Time
Site:	07007004
Issue date:	03/17/2021
From:	Karen Fitch; American Centrifuge Operating, Centrus Energy Corp
To:	John Lubinski; Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML21085A523	List: ML21085A524;
References
	ACO 21-0011, EPID L-2020-LLA-0085
	Download: ML21085A524 (79)
	v • d • e

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. Security- Related In.formation - Withhold Under 10 CFR 2.390

Export Co_ntrolled Information, Official Use Only, and Proprietary Information

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Fueling the Future

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of Nvclear Power t' March 17, 2021 1ACO 21-0011 A TIN: Document Control Desk Mr. Jobn W.- Lubinski, Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington,.DC_ 20555-0001 * .

Ameri~n Centrifuge Plant; Docket Number 70-7004; License Nmµber SNM-2011.

Supplemental Proposed *changes for AmeriC:3n Centrifuge *Operating, LLC's License Amendment Request (EQ.terprise Project Identifi_cation Number: L-2020-LLA-0085)

INFORMATION TRANSMITI'ED HEREWITH IS PROTECTED FROM PUBLIC DISCLOSURE AS CONFIDENTIAL COMMERCIAL 0~ FINANCIAL INFORMATION, AND/OR TRADE SECRETS PURSUANT TO 10 CFR2.390 AND 9.17(a)(4) .

.. .AND .

INFORMATION TRANSMI'ITEO HEREWITH*IS**pROTECTED FROM DISCLOSURE PURSUANT TO 10 CFR PART 81.0

Dear Mr. Lubinski:

American. Centrifuge Operating,* LLC (ACO) hereby* submits for U:S*. Nuclear Regulatory Commission's review and approval, supplement:µ proposed changes to LA-3605-0001, License Application for the American Centrifuge f lant, LA-3605-0003, Integrated Safety Analysis (ISA)

Swnmary fof'the American Centrifuge Plant; and LA-3605-0003A, Addendum I ofthe Integrateµ Safety Analysis (ISA) Summary for the American Centrifuge Plant - HALEU Demonstration.

~nclosure

1 provides the detailed description of supplemental proposed ,changes. Enclosure 2

provides supplemental pr<;>posed changes to LA-3605-0001.' Enclosure 3 provides supplemental proposed changes to Appendix F ofLA-3605.:.0001. *Enclosure*4 provides supplemental proposed changes to LA-3605-0003. Enclosure 5 provides the supplemental proposed* changes to LA-3605-0003A. Supplemental proposed changes from the. previously submitted documents (Referenc6 1, 2,-3, and 4) are noted in ~lue.highlighted track changes with revision.bars in the right-hand margin .

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Docnm~nt/matter h'a~~mitted contains Security- Related Informati~n - Withhold Under 10 CFR 2.390 Export Controlled Informati~n, Official Vse Only, and Proprietary Information * , j L!.. 5 Z, Q.

When separa~ from enclosures, this cover letter is uncontrolled. fl),....,/

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American Centnfuge Operatmg, LLC 3930 US. Route 23 South.- P.O Box 628 *.#Jt-f~S Piketon, OH 45661

Security- Related Information - Withhol.d tJnder 10 CFR 2.390- - - - -~- ---

Export Controlled Information, Official Use Only, and Proprietary Information Mr. John W. Lubinski March 17, 2021 ACO 21-0011, Page 2 Enclosmes 3, 4, and 5 contain Proprietary Information and ACO requests these enclosures be withheld from public disclosure pursuant to 10 Code ofFederal Regulation (CFR) 2.3 90(a)(4). An affidavit required by 10 CFR 2.390(b)(1Xii) is provided as ~nclosure 6. In accordance with the 1

guidance provided by the U.S. Department 'of Energy (DOE), Enclosure 3 also contains Official Use Only information. Enclosures 4 and 5 also contain Security-Related Information; therefore, ACO requests these enclosures be withheld from public disclosure pursuant to 10 CFR 2.390(d)(l).

Enclosmes:4 and 5 have also been determined, in accordance with the guidance provided by the DOE, to contain Export Co:titrolled Information and must be protected from disclosure per the requirements of 10 CFR Part 810.

If you have any questions regarding this matter, please contact me at (740) 897-3859.

Sincerely,

~£$M- l

- Kelly L. Fitch Regulatory Manager Enclosmes: As Stated

References:

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1. ACO 20-0010 from K. Wiehle to J. Lubip.ski (NRC) regarding License Amendment Request for American Centrifuge Operating, LLC's License Application and Supporting Documents for the American Centrifuge Plant, dated April 22, 2020

2. ACO 20-0036 from K. Wiehle to J. Lubinski (NRC) regarding Responses to Requests for Additional Information Related for American Centrifuge Operating, ~LC's License Amendment Request (Enterprise Project Identification Number:. L-2020-LLA-0085), dated

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October 14, 2020

3. lco 20-0039 from K. Wiehle to J. Lubinski (NRC) regarding ,Proposed Changes for Additional Information Related for* American Centrifuge Operating, LLC's License Amendment Request (Enterprise Project Identification Number: L-2020-LLA-0085), dated October 19, 2020

4. ACO 20-0051 from K. Wiehle-Fitch to J. Lubinski (NRC) regarding Supplemental Proposed Changes for American Centrifuge Operating, LLC's License Amendment Request

, (Enterprise Project Identification Number: L-2020-LLA-0085), dated*December 17, 2020

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Docupient/matter transmitted contains Security- Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information, Official Use Only, and ~prietary Information c., _

When separated from enclosu,res, this cover letter is uncontrolled.

Security - Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information, Official "1/se Only, and Proprietary Information Mr. John W. Lubinski March 17, 2021 ACO 21-0011, Page 3 cc (without enclosures, unless* otherwise noted):

Y. Paraz, NRC HQ (Enclosures)

S. Greene, DOE S. Harlow, DOE NE J. Hutson, (CONIR), DOE NE M. McCune, DOE J. Munson, NRC. Region II N. Pitoniak, NRC Region II L. Pitts, NRC Region II (Enclosures)

K. Shears, DOE A. Smith, NRC HQ E. St Clair, (CONIR), DOE NE R Womack, NRC Region II T. Vukovinsky, NRC Region II Document/matter transmitted contains Security - Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information, Official Use Only, and Proprietary Information *i When separated from enclosures, this cover letter is uncontrolled.

Enclosure 6 of ACO 21-0011 Affidavit Information Contained Within Does Not Contain Export Controlled Information Revtewmg Offictal #1014 Date: 03/17/2021

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AFFIDAVIT OF LARRY B. CUILIP SUPPORTING APPLICATION TO WITHHOLD FROM PUBLIC DISCLOSURE CERTAIN INFORMATION PROVIDED TO NRC IN LETIERACO 21-0011 I, Larry B. Cutlip, of American Centrifuge Operating, LLC (ACO), havin.g been duly sworn, ,

do herby affirm and state:

1. I have been authorized by ACO to (a) review the information owned by ACO which is referenced herein relating to ACO's supplemental proposed changes based upon the latest changes in the High Assay Low Enriched Uranium (HALEU) Demonstration Program process design and Nuclear Criticality Safety Evaluations as the described in letter ACO 21-0011, which ACO seeks to have withheld from public disclosure pursuant to section 147 of the Atomic Energy Act (AEA),

as amended, 42 U.S.C. § 2167, and 10 CFR 2.390(a)(4), and 9.17(a)(4), and (b) apply for the withholding of such information from public disclosure by the Nuclear Regulatory Commission (NRC) on behalf of ACO.

2. Consistent with the provisions of 10 CFR 2.390(b)(4) of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

1. The information sought to be withheld from public disclosure is owned and has been held in confidence by ACO.

11. The information is of a type customarily held in confidence by ACO and not customarily disclosed to the public. ACO has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute ACO policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of

several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where presentation of its use by any of ACO's competitors without license from ACO constitutes a competitive economic advantage over other companies.

b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.),, the application of which data secures a ~mpetitive economic advantage (e.g., by optimization or improved marketability).

c) Its use by a competitor would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of ACO, its customers or suppliers.

e) It reveals aspects of past, present, or future ACO or customer funded development plans

,, and programs of potential commercial value to ACO.

f) It contains patentable ideas, for which patent protection may be desirable.

g) It reveals information concerning the terms and conditions, work performed, administration, performance under or extension of contracts with its customers or suppliers.

iii. There are sound policy reasons behind the ACO system which include the following:

a) The use of such information by ACO gives ACO a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the ACO competitive position.

b) It is information, which is marketable in many ways. The extent, to which such

information is available to competitors diminishes ACO's ability to sell products and services involving the use of the information.

c) Use by our competitors would put ACO at a competitive disadvantage by reducing their expenditure of resources at ACO expense.

d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components or proprietary information, any one component may be the key to the entire puzzle, thereby depriving ACO of a competitive advantage.

e) Unrestricted disclosure would jeopardize the position of prominence of ACO in the world market, and thereby give a market advantage to the competition of those countries.

f) The ACO capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

iv. The information is being transmitted to th9 Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

v. The information sought to be protected is not available in public sources or ayailable information has not been previously employed in the same original manner or methe>4 to the best of our knowledge and belief.

3. The proprietary information sought to be withheld is contained within Enclosures 3, 4, and 5 of letter ACO 21-0011. Enclosure 3 provides supplemental proposed changes to Appendix F of LA-3605-0001, License Application for the American Centrifuge Plant. Enclosure 4 provides supplemental proposed chap_ges to LA-3605-0003, Integrated SafetyAnalysis (ISA) Swnmary for the American Centrifuge Plant. Enclosure 5 provides supplemental proposed changes to LA-3605-0003A, Addendum 1 of the Integrated Safety Analysis (ISA) Summary for the American Centrifuge Plant- HALEU Demonstration. These enclosures provide detailed de~riptions and

diagrams related to the deployment of ACO's HALED enrichment process specifics unique to the American Centrifuge Plant; therefore, determined to be proprietary.

Public disclosure of this proprietary information is likely to cause substantial hann to the competitive position of ACO because it may enhance the ability of competitors to position and provide similar products. Moreover, disclosure of this information may provide insights into the design of ACO's American Centrifuge technology, including structures, systems, and components categorized as Export Controlled Information.

This information is part of that which will enable ACO to:

Identify the detailed process flows of the various structures, systems, and components used within the HALED Demonstration cascade and the future American Centrifuge Plant;

Analyze the haz.ards evaluations associated with event sequences; and

To continue future operation of the American Centrifuge Plant.

Further, this information has substantial commercial value as follows:

The development of the information described in part is the result of applying many hundreds of person-hours and the expenditure of thousands of dollars on design and analysis activities to achieve the information that is sought to be withheld; and

In order for a competitor of ACO to duplicate the information sought to be withheld, a similar process would have to be undertaken and a significant effort ap.d resources would have to be expended.

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Further the deponent sayeth not

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Larry B. Cutlip, having been duly sworn, hereby confirms that I am the Senior Vice President, I

Field ~ o n s of American Centrifuge Operating, LLC, that I am authorized on behalf of ACO to I

review the information, attached hereto and to. sign and file with the, u.s~ Nuclear Regulatory

_)

Commission this affidavit and the attachments hereto, and that the statements made and matters set forth herein are true and correct to the best of my knowledge, infonnation, end belief.

'1 On this 1~ day of March 2021, Larry B. Cutlip personally appeared before me, is known by me to be the person whose name is subscribed to within the instrument and -acknowledged that be execµted the same for the purposes therein contained.

In witness'hereof I hereunto set my band and official ~-

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of State Tcnnesffi: Notary Pµblic Anderson County

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My comlliission expires Octooer 26, 2024 I

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r Enclosure 1 of ACO 21-0011 Detailed Description of Supplemental Proposed Changes Information Contained Within Does Not Contain Export Controlled InformatiQn R.eviewmg Offic1al #1014 Date. 03/09/2021

Enclosure 1 ACO 21-0011 Detailed Description of Supplemental Proposed Changes The primary driver for the proposed change to LA-3605-0001 , License Application for the American Centrifuge Plant, is to incorporate a commitment requested during the U.S . Nuclear Regulatory Commission (NRC) review. Chapter 5.0 was revised to include the following license commitment:

Within 30 days of making any non-administrative changes to the validation report, the Licensee shall provide the Commission with a summary of changes and shall provide the revised validation report upon request. The licensee may not implement changes to reduce the margin of subcriticality for safety (i.e., factors or methods that would adversely affect the Upper Subcritical Limit) without NRC approval of the change.

In addition to the above change, the Definitions section was revised to add the term "Frequencies" and define various terms associated with frequencies (e.g., quarterly, semiannually). Frequencies are used in the License Application and supporting documents. For example, Appendix F requires that verification be performed semiannually. Including the definition for the different time intervals provides a common definition for use at the American Centrifuge Plant. The other changes to the License Application are considered clarifications or corrections.

In addition to the changes to the License Application, this submittal also includes supplemental proposed changes to the Integrated Safety Analysis Summary documents. The changes are being made in conjunction with the latest changes in the High Assay Low Enriched Uranium process design and revisions to corresponding the Nuclear Criticality Safety Evaluations.

The detailed description, justification, and American Centrifuge Operating, LLC ' s significance determination submitted by ACO 20-0010, dated April 22, 2020, remain unchanged. Based upon the 10 Code of Federal Regulation 70.32 and 70.72 evaluations, not all proposed changes depicted warrant the NRC ' s review and approval; however, are provided for completeness to assist in the review efforts.

Corresponding supplemental proposed changes to LA-3605-0001 are contained within Enclosures 2 and 3. Corresponding supplemental proposed changes to LA-3605-0003 , Integrated Safety Analysis (ISA) Summary for the American Centrifuge Plant, are contained within Enclosure 4. Corresponding supplemental proposed changes to LA-3605-0003A, Addendum 1 ofthe Integrated Safety Analysis (ISA) Summary for the American Centrifuge Plant - HALEU Demonstration, are contained within . Supplemental proposed changes are identified by the following method:

- Identifies text to be removed

- Identifies text to be added

Enclosure 2 of ACO 21-0011 Supplemental Proposed Changes to LA-3605-0001, License Application/or the American Centrifuge Plant Information Contained Within Does Not Contain Export Controlled Information Reviewing Official : # 1014 Date: 03/09/202 1

LA-3605-0001 License Application for the American Centrifuge Plant in Piketon, Ohio The American Centrifuge'"

Proposed ChangeRe¥isioe 54 Docket No. 70-7004 Information Contained Within Does Not Contain Export Controlled Information Reviewing Official: # 101 4 Date: 03/09/202 1

License Application for the American Centrifuge Plant Proposed Change 2021 Blank Page

License Application for the American Centrifuge Plant Proposed Change 202 I LA-3605-0001 LICENSE APPLICATION for the American Centrifuge Plant in Piketon, Ohio Docket No. 70-7004 Proposed ChangeRe¥ision 54

License Application for the ".American Centrifuge Plant Proposed Change 2021 Blank.Page

License Application for the ~erican Centrifuge Plant Proposed Change 2021 UPDATED LIST OF EFFECTIVE PAGES Rev1S1on 0- 1o'CFR 1045 revt~w completed by L Sparks on 07/29/04, Export Controlled Information review completed by R Conell on 07/30/04 Rev1S1on l - l 0 CFR 1045 review, completed by L Sparlcs on 03/04/05, Export Controlled Information review completed by R Conell on 03/10/05 Rev1S1on 2- IP CFR 1045 review completed by J Weidner on 04/29/05; Export Controlled Infonnat10n review completed by'R Conell on 04/29/05 Revision 3-10 CFR 1045 rev1~ completed by J Weidner on 05/23/05; Export Controlled Infonnat:Jon review completed by,R. Conell on, 05/23/05

- Revision 4- IO CFR 1045 review completed by R Conell on 06/16/05; Export Controlled InformatJ.on review completed by D Hupp on 06/16/05 Rev1S1on 5- 10 CFR 1045 review completed by J Weidner on 06/21/05; Export Controlled lnformat:J.on revi\m' completed by D Hupp on 06121/05 Revision 6- 10 CFR 1045 review ~mpleted by J Weidner on 08/30/05, Export Controlled lnformatJ.on review completed by D Hupp on 08/30/05 Revision 7- 10 CFR 1045 review completed by J Weidner on 09/02/05; Export Controlled Informat:J.on review completed by R Conell on 09/02/05 Revision 8- JO CFR 1045'review completed by J Weidner on 09(27/05, Export Controlled Infonnat:Jon review completed by D Hupp on 09/13/05 Revision 9- JO CFR 1045 review completed by J Weidner on 10/05/05; Export Controlled Infonnat:ion review completed by D Hupp on 10/05/05.

RevlSlon 10-10 CFR 1045 review completed by J Weidner on 11/04/05, Export Controlled Information review completed by D Hupp on 11/04/05 Rev1S1on 11-10 CFR 1045 review completed by J Weidner on 11/17/05; Export Controlled Information review completed by D Hupp ,on I 1/14/05 RevISion 12- IO CFR 1045 review completed by J Weidner on 11/28/05, Export Controlled Information review completed by D Hupp on 11122/05 Rev1S1on 13-10 CFR 1045 review completed by J Weidner on 12/02./05, Export Controlled Information re".)ew completed by D Hupp on 12/02/05 RevISion 14-10 CFR 1045 review completed by J Weidner on 03/l 710fJ, Export Controlled Informlll:Ion review completed by D Hupp on 03/17/06 Revision 15-10 CFR 1045 review completed by R Conell on 06/01/06, Export Controlled InfonrurtJon review completed by G. Peed on 06/01/06 Rev1S1on 16-10 CFR 1045 and the ~ r t Coµtrolled Information reviews w~ completed by R Conell on 08/11/06 RevISion 17- 10 CFR 1045 and the Export Controlled Information reviews were completed by G Peed on 08/30/06

. RevIS1on 18-10 CFR 1045 and the Export Controlled Information reviews were completed by R Coriell on 09/06/06 Rev1S1on 19 - 10 CFR I 045 and the Export Controlled Infonnlltion reviews were completed by R Coriell on 06/'12/07 Revis10n 20-10 CFR 1045 and the Export Controlled InformatJon reviews~ completed by R Conell on 10/09/07 Rev1S1on 21- IO CFR 1045 and the Export Controlled Information reviews were completed by O Peed on 01/11/08 Rev1S1on 22 - l 0 CFR l 045 and the Export Controlled Information reviews were completed by O Peed on 0 1/25/08 Rev1S1on 23 - l 0 CFR I 045 and the Export Controlled Information reviews were completed by O P.eed on 03/04/08 Rev1S10n 24 - l 0 CFR l 045 end the Export Controlled InformatJon revJCWS were completed by M Basham on 06/05/08.

Rev1S1on 25 - l 0 CFR l 045 and the Export Controlled Information reviews were completed 'by M Basham on 09/29/08 RevISion 26-10 CFR 1045 and the Export Controlled InformatJon reviews were completed by R S LykowskI on 11/24/08 RevlSlon 27- 10 CFR 1045 and the Export Controlled Inforrnlll:Ion reviews were completed by R S Lykowsla on 1/14/09.

Rev1S1on 28- 10 CFR 1045 and the Export Controlled lnform.at!on reviews were completed by R S Lykowsk1 on 1/27/09 Revis10n 29-10 CFR 1045 and the Export Controlled Information reviews were completed by R S LyJrnwski on 10/15/09 Revision 30- 10 CFR 1045 and the Export Controlled Information revJCWS were completed by R S Lykowsla on 1/8/10 Rev1S1on 31 - IO CFR l 045 and the Export Controlled Information reviews were completed by R S Lykowslaon 2/8/10 Rev1S1on 32 - l 0 CFR I 045 and the Export Controlled Inforrnlll:Ion reviews were completed by R S. Lykowskl on 4/20/10

Rev1S1on 33-10 CFR 1045 and the Export Controlled Inforrnlll:Ion reviews were completed by R. S Lykowskl on 5/25/10 RevIS1on 34-10 CFR 1045 and the Export Controlled Information-reviews were completed by R S Lykowsla on 7/23/10 Rev1S1on 35- 10 CFR 1045 and the Export Controlled Information reviews were completed by R S Lykowsla on 8/17/10 Rev1S1on 36 - l 0 CFR l 045 and the Export Controlled Information reviews were completed by R S Lykowsla on I 0/6/10.

Revision 37-10 CFR 1045 ond the Export Controlled Infoimanon reviews wee~ completed by R S. Lykowsla on 11/3/10 Revision 38 - Reviewed and detenruned to be UNCLASSIFlED. Denvatrve Cla'!Slfier RS LykowskL Sensrttve Information reviews completed and approved for public release by RS Lykowskl on 2/18/11 Rev1S1on 39 - Reviewed and determmed to be UNCLASSIFlED Denvatlve Classifier RS Lykowsla Sensrtrve Informlll:Ion reviews compJeted and approved for public release by R.S Lykowski on 3/4/1 L .

Rev1S1on 40 - 'Reviewed and determmed to be UNCLASSIFlED Derivatrve Classdier RS Lykowskl Sensrtrve rnformatJon reviews completed and approved for public release by RS Lykowslo on4/19/2011 Rev1S1on 41 - Reviewed and detenruned to be UNCLASSIFlED Denvatrve Classdier RS Lykowsla Sensitive Information reviews completed and approved for public release by RS Lykowski on 9/2/2011 Revision 42 - Reviewed and detemuned to be,UNCLASSIFIED Denvative*CIMS1fier RS Lyko~ Scnsrttve Information reviews completed and approved for pubhc release by RS Lykowskl on 10/20/2011 Revision 43 - Reviewed and detenruned to be.UNCLASSIFIED Denvative Classdier RS Lykowsla Sensrtrve InfonnatJon reviews completed and approved for public release by RS Lykowskt'on 8/27/2012 Rev1S1on 44 - Reviewed and deternuned to be UNCLASSIFlED Denv,etlve Clossrlier RS Lykowskl Scnsrtive Information reviews completed and approved for public release by RS Lykowsla on 2f7/2013 Rev1S1on 45 - Reviewed and detenmned to be UNCLASSIFlED. Denvatrve Classifier RS Lykowski Sensrtrve mformatJon reviews completed and approved for public release by R.S Lykowsla on 2/20/2013 RevISion 46 - Reviewed and determmed to be UNCLASSIFlED Denvatrve Classifier RS. Lykowskl Sensitive Informat:J.on reviews completed and approved for pubhc release by RS Lykowski on 8/20/2013.

Revtsion 47 ~ Reviewed and detennmed to be UNCLASSIFIED ~ative Oa'!Sifier RS Lykowsla Sensrtlve rnfof1Il:1111on reviews completed and approved for pubhc release by RS Lykowsla on l I/21f2013.

RevlSlon 48 - Reviewed and deterinmed to be UNCLASSIFIED Denvatlve Classifier RS Lykowskl Sensitive Information reviews completed and approved for pubhc release by RS. Lykowskl on 12/16/2013

l Revision' 49 - Reviewed and dctermmed to be UNCLASSIFIED. Denvlll:Ive Oosslfier RS Lykowslo Sei),srtJve Informauon reviews completed and approved for public release by RS Lykowski on 2/13f2014 Revision 50 - Reviewed and clet=med to be UNCLASSIFIED. Denvative Classdier RS Lykowsla Sensitive Information reviews completed and approved for pubhc release by RS Lykows1a on 7/28/2015 Revision 51 - Reviewed and determined to be UNCLASSIFlED , Denvetlve Cla<isrlier RS Lykowski Sensrttve Information reviews completed and approved for public release by RS. Lykowski on 12/2/2015 Revision 52 - Reviewed and deternuned to be UNCLASSIFIED Denvatrve Classifier #4769 Sensrtlve Informat:J.on revtews completed and approved for pubhc release'by Reviewer #1014 on 6/17/2016 Revision 53 - Reviewed and c!et=ed to be UNCLASSIFIED Denvative Classifier #4769 Sensrtlve Information reviews completed and approved for pubhc release by Reviewer # l 0 14 on 03/09(2017 ULOEP-1

License Application for the American Centrifuge Plant Proposed Change 2021 Proposed Change - Classification review completed by Derivative Classifier #4769 on April 22, 2020 and the Controlled Unclassified Information (e.g. ,

ECI review com leted b Reviewer# IO 14 on A ril 22 2020.

ULOEP-2

- License Application [or the American Centrifuge Plant Proposed Change 2021 Updated List of Effective Pages fi!2e ~!!mber Revision Number f!!2~ N!IIDher *. Re~i2n N!!!!!ber Co:verPages 53 1-16 42 ULOEP-1 53 1-17 42 ULOEP-2 53 1-18 42 ULOEP-3 50 1-19 42 ULOEP-4 52 1-20 42 ULOEP-5 52 1-21 42 ULOEP-6 53 1-22 42 ULOEP-7 53 1-23 42 ULOEP-8 Proposed Change 1-24 42 Table or Contents 1-25 42 45 1-26 42 ii 50 1-27 . 42 iii ' 50 1-28 42 iv ~o

1-29 42 V 22 1-30 42 vi 32 1-31 42 vii : 32 1-32 42 viii 32 1-33 42 ix 41 1-34 42

'X 22 1-35 42

xi 22 1-36 42 xii 45 1'-37 42.

xiii 42 1-38 42 xiv 45 1-39 42

xv 50 H-0 42, xvi 32 1-41 42 \

xvii 15 1-42 42 xviii 52 1-43 42 XIX 15 1-44

42 xx 15 1-45 42 xxi'

15 1-46 42 I

xxii 15 1-47 '42 xx.iii 15 1-48 '42.

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XXIV, 15 1-49 42 1-50 42 Executive Summary 1-51 42 1 45 1-52 42 45 1-53 I 42 2

1-54 42 Chapter 1.0 1-,55 42 1-1

32 1-56 .47 1-2 32 J 1.:.57 -45 1-3 32 1-58 45

'1-4 32, 1-59 45 1-5 32 1-60 45

_J 1-6 32 1-61 45 1-7 32 1-62 45 1-8 I '38 l-63 45

( 45.

1-9 32 1-64 J

1-10 42 1-65 45 1-11 32' 1 -45 1-12 ,32 1-67 46 1-13 45 1'--68 . 46 1-14 42 1-69 45 1-15 42 1-70 45 ULOEP-3

License Ape_licaiion f2r the Amerkan Ce~if_ug_e-f>kmt Pr(Y(!Q_Sed.Chailg_e 2021 Updated List of Effective Pages fau;e ~!Imber Revisi!!!! IS:nmber Pa2eNnmber' Revision Nnmbec 1-71 ( 45 1-126 :45 1-72 45 1-127 45 1-73 _* 45 1-128 45 1-74 45 1-129 45 1-75 45 1-130 45 1-76 45 1-131 45 1-77' 47 1-132 45 1-78 *45 f-133 45 1-79 45 1-134 *45 1-80 45 1-135 45 1-81 45 1-136 45 1-82 4'5 1-137 45 1-83 45 1-138 . 45 1-84 45 1-139 48 1-85 45 1-1:40 *48 1-86 45

Chapter 2.0 1-87 45 . 2-1 50 1-88 45 2-2 50 1-89 45 2-3 *50 1-90. 45 2-4 50 1-91 45 2-5 *~ 50 1-92 45 ,2-6 _ ,50 1-93 45 2-7 50 1-94 45 2-8 50 1-95 45 2-9 50 1-96 45 2-10 50 1~97 48 2-11 50 1-98 48 2-12 50 1-99 45 2-13 50 1-lQ0 45 2-14 50 1-101 45 2-15 50 1-102 45 2-16 50 1-103 47 1-104 45-1-105 45 1-106 45 Chapter3.0 1-107 45 3-1 28 1-108 . 45 3-2 28 1-109 45 3-3 28 1-110 45 3-4 40' 1-111 45 3-5 40 1-.112 45 3-6 32 1-113_ 45 3-7 40 1-114 45 ,3-8 ,15 1-115 45 3-9 15 1-116 45 3-10 15 1-117 45 3-11 15 1-118 45 3-12 15

. 1-119 45 3-13 40 1-120 4? 3-14 40 1-121 45 3-15 40 1-122 45 3-16 32 1-123 . 45 3-17 ,15 1-124 45 *3-18 15 1-125 45 3-19 :32 ULOEP-4

License Application for th/A~rican-Centnfage Plan{- - - 0

Propo;ed Change i02J Updated List of Effective Pages PageNgmber . Revision Number Page Number Revision Number 3-20 40 5-15 16 3-21 40 5-16 16 3-22 40 5-17 16 3-23 40 5-18 16 3-24 40 5-19 16 3-25 15 5-20 16 3-26 15 5-21 16 3-27 15 5-22 16 3-28 32 5-23 15 3-29 48 5-24 15 3-30 48 3-31 28 Chapter6.0 3-32 48 . 6-1 22 3-33 48 .6-2 32

3-34 18 6-3 28 6-4 43 Cbapter4.0 6-5 r 28 4-1 50 6-6 22 4--2 25

6-7 52 4--3 50 6-8 52 4-4 14* 6-9 28 4-5 14 6--10 f4 4--6 ' 14 Chapter7.0 4-7 14 7-1 15 4-8 14 7-2 15 4--9 14 7-3 22 4-10* 31 7-4 39 4-11 22 7-5 -39 4-12 14 7-6 32 4-13 14 7-7 15 4-14 14 7-8 39

\ 4-15 14 7-9 19

4-16 14 7-10 15 4-17: 14

7-11 32 4-18 14 7-12 39 4-19 \_' 28 7-13 39 4-20 14 7-14 22 4-21 14 7 15 4-22 14 7-16 15

. 4-23 14 7-17 15 4-24* 14 7-18 39 Chapter8.0 Chapter5.0 8-1 28 5-1 28 8-2 14 5-2 22' 8-3 Proposed Change 5-3 22 8-4 Proposed Change 5-4 15 5-5 22 5-6 15 5-7 22 Chapter9.0 5-8 28 9-1 22 5-9 *16 9-2 32 5-10 16 9-3 32 5-11 22 . 9-4 32 5-12 16 9-5 32 5-13 22 9-6 32 5-14 16 9-7 32 ULOEP-5

License APPiication for the American Centrifuge Plant Pr'oposed Change 2021 Updated List of Effective Pages Page Number* Revisi2n Numbet Pai:;eNumber Revision Number 9-8 32 Chapter 10.0 9-9 32 10-1' 28 9-10 38 10-2 32 9-11 32 10-3 14 9-12 32 10-4 14 9-13 32 10~5 32 9-14 32

10-6 28 9-15 32 10-7 42 9-16 32 10-8 28 9-17 32 10-9 42 9al8 51 10-10 32 9-19 32 10-11 28 9~20 32 10-12 28.

.9-21 32 10-13 14 9-22 32 10-14 32 9-23 38 l 10-15 32 9-24 32 10-16 32 )

9-25 32 10-17 32 9-26 22 10-18 32 9-27 32 10-19 28 9-28 32 10-20 16 9-29 14 10-21 32 9-30 14 10-22 32 9-31 14 9-32 14 Chapter 11.0 9-33 14 11-1 28 9-34 14 . 11-2 34

'9-35 14 11~3 22 9-36 14 11-4 29 9-37 5-1 11-5 22 9-3,8 51 11-6 41 9-39 51 11-7 41 9-40 51 11-8 41 9-41 51. 11-9 41 9-42 51 11-10 41 9-43 51 11 41 9-44

51' 11-12* 41

9-45 51 11-13 41 9-46 51 11-14 41 9-47 51 11-15 25

9-48 51 11-16 25 9-49

51 11-17 25 9-50 51 11-18 43 9-51 51 11-19 22 9-52 51 11-20 .22 9-53 51

11-21 22 9-54 51 fl-22 52 9-55 32 11-23 22 9-56 38 11-24 22 9-57 14 11-25 22 9-58 14 11-26. 22 9-59 14 11-27 22 9-60 14 11-28 *22 9-61 14 11-29 ,22 9-62 14 11-30 '15 11-31 15 ULOEP-6

. -r-.

. License Aep.Jicationf2r the American-Centrifuge Plant Proposed Change 2021 Updated List of Effective Pages Page ~!!m!!er Revision Number f!!ge ~!!!!!bet Revision Number 11-32 19 B-14 32 11-33 40

B-15 38 11-34 40 B-16 32 11-35 40 B-17 32 11-36 52 B-18 32 11-37 40 B-19 32 11-38 15 : B-20 53 11-39 40 11-40 40 AppendixC 11-41 15 C-1 32 11 40 C-2 32 11-43 40 C-3 32 11-44 40 C-4 32 11-45 40. C-5 32 11-46 45 C-6 32 11-47 24 11-48 15

AppendixD

/ 11-49. 15 D-1 . Proposed Change

11-50 44 D-2 Proposed Change 11-51 44 D-3 Proposed Change 11-52. 44 D-4, Proposed Change 11-53 25 .D-5 Proposed Change r 11-54 15 D-6 Proposed Chang_e 11-55 15 D-7 Proposed Change 11-56 44 D-8 Proposed Change 11-57 15 D-9 Proposed Change 11-58 15

0:10

Proposed Change 11-59 15 11-60 15 AppendixE 11-61' 37, E-1 38 11-62 15 E-2 32 E-3 38 Appendix A E-4 32 A-1 . 32 E-5 32 A-2 32 E-6 32 A-3 32 E-:7 32 A-4 32 E-8 32 A-5 32 E-9 32 A-6 32 E~IO 32 E-11 32 AppendixB E-12 32 B-1 53 E-13 32 B-2 32 E-14 32 B-3 53 E-15 38 B-4 53 E-16 32 B-5 40 E-17 32 B-6 49 E-18 32 B-7 32

E-19 32 B-8 32 E-20 32 B-9 32 E-21 32 B-10 32 E-22 32 B-11 32 ) E-23 32 B-12 32 E-24 32 B-13

32 E-25 32 ULOEP-7

License Application fot the American Centrifuge Plant Proposed Change 2021 Updated List of Effective Pages Page Number Revision Number Page Number Revision Number E-26 32, E-27 32 E-28 32 E-29 32 E-30

38 AppendixF F-1 Proposed Change F-2 Proposed Change F-3 Proposed Change F-4 Proposed Change F~S 'Proposed Change F-6

Proposed Ch~ge

_/

( '

ULOEP-8

License Application for the American Centrifuge Plant Proposed Change 2021 TABLE OF CONTENTS Acronyms and Abbreviations ................................................................... ..... ........................ ... xvii Definitions ....................................................................................................... ........................... xxi Chemicals and Units of Measure .... .. ....................................................................................... xxiii Executive Summary ......... .. ........ .. .............. .. ............. ... ........ ........... ................. .............................. 1 1.0 GENERAL INFORMATION .................................................. .. .. ... ...... ..................... ... 1-1 1.1 Plant and Process Description .................................... .. ............................. ......... 1-2 1.1.1 Site Boundary .... .. ......... .. ... .. .................................... ........................... 1--;!-}

1.1.2 Plant Layout .............................. ........ ..... .................... ............................ 1-3 1.1.3 Primary Facilities Description ............................................................... 1-3

1. 1.4 Secondary Facilities Description .. ...................................... ............. 1--1-011 1.1. 5 Process Description .................... .. .................... ... .. ................. ...... .... 1-l-4U 1.1.6 Hazardous Material Storage .................... ................ ... .. .... .................... 1-29 1.1.7 Roadways ... .............................................. ................................... ......... 1-30 1.1.8 Tr!1/4ftsitioa from. the Lead Coseode Dem.oastrotion Foeility toPhased Modular Expansion Plan for the American Centrifuge Plant.. ............ 1-30 1.1.9 Material of Construction ........ ................. .............. ........................... 1~ 32 1.1 .10 Use of Lubricants ..................... .. ............. .................................... ..... 1-~ 32 1.2 Institutional Information ............................................................. ... ......... ..... 1-1/457 1.2.1 Corporate Identity ......................................................................... ... 1-1/457 1.2.2 Financial Qualifications ....................................................................... 1-59 1.2.3 Type, Quantity, and Form of Licensed Material.. .... ........................ .... 1-61 1.2.4 Authorized Uses .. ...................... .. ........ .. ...................... ..... ........... ........ 1-61 1.2.5 Special Exemptions or Special Authorizations ............ .................... 1-6+62

License Application for the American Centrifuge Plant Proposed Change 2021 1.2.6 Security of Classified Information ......................................... ..... .... . 1-6169 1.2.7 Security of Special Nuclear Material of Low Strategic Significance and Moderate Strategic Significance ................................................. ... .. 1-6870 1.3 Site Description .... ................. .. .... .. ....................... ......................... .... .. .... .... . 1-1677 1.3 .1 Geography .. .... ............................ .... .. ............... ............... .................. 1-1677 1.3.2 Demographics .. ......... ..... ....... ............................. .. ...... ... ........... ....... . 1-1677 1.3.3 Meteorology .................................................................. ............ .. ..... 1-1880 1.3 .4 Surface Hydrology .... ................................................................ ... .... 1--&Gfil.

1.3 .5 Subsurface Hydrology ... ..................... ............... ....... ... ....... .... .. ....... 1-&8-89 1.3 .6 Geology and Seismology ........ .... ............................ .......... ............... 1--9;.94 1.4 Applicable Codes and, Standards, and Regulatory Guidance ... ..... .......... 1-H-4115 1.4.1 American National Standards Institute/American Nuclear Society ......... 1-H4115 1.4.2 American National Standards Institute ........... ................ .... .. .... ... 1-++6117 1.4.3 American National Standards Institute/American Society of Mechanical Engineers ... ............................................................... 1---l-l-& 118 1.4.4 American Society of Mechanical Engineers .. ................ ....... .... ... 1-+l-9119 1.4. 5 American Society for Testing and Materials ............ ................... 1~ 120 1.4.6 National Fire Protection Association ........................................... 1~ 121 1.4.7 Section Reserved For Future UseNuclear Regulatory Commission Guidance .. ... .. ..... ........ ......... .. ...................................................... ...... ....... 1-H4124 1.4.8 Institute of Electrical and Electronics Engineers ......................... 1~ 124 1.4.9 Other Various Codes and, Standards, and Guidance ................... 1-86149 1.5 License Application Regulatory Guidance DocumentsReferences ......... 1-H& 133 1.5.l U.S. Nuclear Regulatory Commission Guidance .......... .................... 1-133 1.5.2 Other Various Guidance Documents ................................................. 1-137 1.6 References ....................... .. ................... ............................................ .............. 1-138 II

License Application for the American Centrifuge Plant Proposed Change 2021 2.0 ORGANIZATION AND ADMINISTRATION ... ..................... ... .......... .. ... ... ..... ... ....... 2-1 2.1 Organizational Commitments, Relationships, Responsibilities, and Authorities .......... ..... ..... ... ... .... .... ............ .... ........ .... ................................ ..... ....... 2-2 2.1.1 Senior Vice President, Field Operations ...................... ... .. ..... ............... . 2-2 2.1.2 General Manager .... .... .. .... ..... .... ............ ........ ... ... .......................... ... ...... 2-3 2.1.3 Director, Quality Assurance ...... .... ...... .. ....... .... .......... ...................... .. 2---1-02.

2.1.4 Director, Engineering, Procurement, and Construction ... .. ... ....... .... 2-l-l-10 2.1.5 Director, Nuclear Safety .............................. ... ...... ............. ...... ........ .. .. 2-10 2.1.6 Director, Engineering ............. ....... .... .............. ............................... ...... 2-10 2.1. 7 Plant Shift Superintendent (Contractor) ............. .... ... .... ... .. ............ .. 2---1-012 2.1.8 Shift Crew Composition [only during operational phases with licensed material] ......... .................. ..... ..... ......... ........... 2---1-012 2.2 Management Controls .......... ..... ...... ......... ........ ....... ...... ....... ....................... . 2--t-; 14 2.2.1 Plant Safety Review Committee ... ...... ............... ..... ... ..... ......... .... .... 2-1412.

2.3 Pre-operational Testing and Initial Start-up ...... .......... .... .... ................. ..... ... 2-1412.

2.3.1 Pre-operational Testing Objectives ........ ................. .. ...... ............. .... 2-1-§-16 2.3 .2 Turnover, Functional, and Initial Start-up Test Program ... .............. 2-1-§-16 2.4 References ........... ..... ... ....... ......................... ...... ............... ..... ..... ... .... ...... ......... 2-16 3.0 INTEGRATED SAFETY ANALYSIS AND INTEGRATED SAFETY ANALYSIS

SUMMARY

................... ......... .... .. ... ...... .................. ... ..... .... .... ................. 3-1 3.1 Safety Program and Integrated Safety Analysis Commitments ......... .... .. .... ...... 3-1 3 .1.1 Process Safety Information .......... .... ........... ... ................. ... .. ..... ............. 3-1 3 .1.2 Integrated Safety Analysis .... ... ..... .... ............. ...... ........................... ....... 3-2 Ill

License Application for the American Centrifuge Plant Proposed Change 2021 3.1.3 Management Measures ................. .......................................... ......... 3-+/-932 3.2 Integrated Safety Analysis Summary ... .......................... ... .. .... ..................... 3-+/-933 3.3 Items Relied on For Safety Boundary Definition ....... .. .... .. .... .. .......... .. .... .. . 3-+/-933 3.4 Seismic Specifications .................................................................. .. .... .. ... .. .. 3-~ 33 3.5 Integrated Safety Analysis Maintenance .................... ............. ............ .. ...... 3--:}-1-34

3. 6 References ... .......................... ............. .......................... .. .... .. .. ....... .. ...... .. .. ... 3--:}-1-35 4.0 RADIATION PROTECTION ................................................ .. ....... ................ .... ......... 4-1 4.1 Radiation Protection Program Implementation .............. .. .. .. .................... ......... 4-1 4.2 As Low As Reasonably Achievable Program ........... ................................ ... ...... 4-1 4.2.1 As Low As Reasonably Achievable Committee ....................... .. ........... 4-1 4.3 Organization and Personnel Qualifications .................................................. ..... 4-3 4.4 Written Procedures ... .. .................. ..................... ........ .... ..................................... 4-4 4.4.1 Procedures .. ..... ............................ ........ .............................................. .. ... 4-4 4.4.2 Radiation Work Permits ................................... ... ........... ... ... .. ......... .. ..... 4-4 4.5 Training ....... .. ...... ..... .... ................... ........................ ...... ... .... ............ .............. .. .. 4-5 4.5.1 Visitor Site Access Orientation ..................... ..... ... ................. ... ............. 4-5 4.5.2 General Employee Radiological Training .. .................................. ......... 4-5 4.5.3 Radiation Worker Training .... ....................................................... ......... 4-5 4.5.4 Health Physics Technician ................................. .... .. ....... .. ..... .... ..... ....... 4-5 4.6 Ventilation and Respiratory Protection Programs ............................................. 4-6 4.6.1 Ventilation .... .... ............................. ........................................... ........ ...... 4-6 4.6.2 Respiratory Protection ............. ..................................... ....... ....... ........... 4-7 4.7 Radiation Surveys and Monitoring Program .................. .. ... ............................ 4-9~

IV

License Application for the American Centrifuge Plant Proposed Change 2021 4.7.1 Surveys ... .................... ..... ........... .............. .................... .............. .......... 4-9li 4.7.2 Personnel Monitoring ............ .... .. ....... .. .... ... ... ... ....... ..... ..... ..... .. ....... .. 4-1-02 4.7.3 External .... ......... ...... .. ..... ... ..... .......... ............. ..... .. ..... .... .......... ...... ...... . 4-10 4.7.4 Internal ........ ........ .... .. ..... ... ...... .......... ..... .... .... ... ... .............. ...... .. ... .... .. . 4-11 4.7.5 Airborne Radioactivity ................ .............. .. ... ... ..... .. .... .. .. .... .. ... ..... .. 4-1-+/-ll 4.8 Additional Program Elements ....... ............. ... ..... .... .... ... ... ....... .. ..... .... ... .... ..... .. 4-16 4.8.1 Posting and Labeling .................................... .. ... .. .. ....... .... .... ... ... .......... 4-16 4.8.2 Contamination Control... ... .. ... .... .... .... ......... .... .. ....... ... .. .... ... ... ... .... ...... 4-16 4.8.3 Radioactive Source Control ..... .. .......... ..... ........ ... .................... ... ........ . 4-18 4.8.4 Radiation Protection Instrumentation ............ ... .. .... ............. ... ........ ..... 4-18 4.8.5 Records and Reports .......... ......... ............ .... .... ... ... ... ... .... .. ..... ...... ........ 4-19

4. 9 References ....... .... .. ... ............. ..... .... ........ ......... ........ ... ......... .............. ............... 4-23 5.0 NUCLEAR CRITICALITY SAFETY ... ...... ........ ............ .... ..... ..... ... ...... ... ... ... .... ...... ........ 5-1 5.1 Management of the Nuclear Criticality Safety Program ... ... ... .. .... ... ... .... ... .... .. 5-+2 5 .1.1 Program Elements .......... ... ....... ......... ....... ... ..... ...... ... ... ..... ...... ... .... ...... 5-+2 5.1.2 Program Objectives ..... .......................... ... ................ ................. ..... ..... ... 5-2 5.2 Organization and Administration .. ...... .... ... ..... ........ ... .... ....... ...... ... ..... ...... ... ...... 5-2-J 5.2.1 Nuclear Criticality Safety Responsibilities ...... ..... .. .... .. .. ...... .. ..... ....... . 5-2-J 5.2.2 Nuclear Criticality Safety Staff Qualifications ................. ... ...... .... .... .... 5-4 5.3 Management Measures ........................... ..... ...... ... ........ ... .. ....... ... ... .... ................. 5-5 5.3 .1 Procedure Requirements ... ....... ....... ........ ...... ... .......... .. ....... ..... ... ...... ..... 5-5 5.3.2 Posting and Labeling Requirements .. ..... .... ......................... .. ....... ..... .. 5 §_

5.3.3 Change Control ........ ..... ... ................ ... .... .. .... ... ..... .... ... ..... ..... ....... .... ..... 5-6 V

License Application for the American Centrifuge Plant Proposed Change 2021 5.3.4 Operation Surveillance and Assessment.. ..... .... .... .... .. .......... ....... ........ .. 5-7 5.4 Methodologies and Technical Practices ..... ..... ...... ..... ... ... ....... ... .... ... ........ ........... 5-9 5.4.1 Adherence to American National Standards Institute/American Nuclear Society Standards .. ....... ... ...... ..... .. ... .... ....... ............. ..... ....... ..... 5-9 5.4.2 Nuclear Criticality Safety EvaluationProeess Evaluation and Approval5-9 5.4.3 Design Philosophy and Review ....... ..... .... ...... ..... ... ...... ... ........... ......... 5-12 5.4.4 Criticality Accident Alarm System Coverage ........ ....... ........ .......... ..... 5-12 5.4.5 Technical Practices ...... ........ ..... ....... .......... ..... ........... ......................... . 5-14 5.5 References ....... ....... .... .............. ............. ....... .... ... .. ........ ... ... ... ... ........... ...... .... 5-2J .Q 6.0 CHEMICAL PROCESS SAFETY ..... ...... ..... ................. ............... ... ........ ..... ........ .. ..... 6-1 6.1 Process Chemical Risk and Accident Sequences ... .......... .. .. ... ..... ... .. ....... ..... ..... 6-1 6.2 Items Relied on for Safety and Management Measures ................ ..... ....... .. ...... 6-2 6.2.1 Items Relied on for Safety ....... ..... .... .... ..... .. ..... ..... ...... ..... .. ........... ...... 6-+/-1 6.2.2 Management Measures ... .. ....... .... ...... ..... ..... ..... .... ... ... ... ..... .... .. ..... ........ 6-3 6.3 Requirements for New Buildings/Facilities or New Processes at Existing Facilities .. ... ...... ..... .... .... .......... ............... .......... ..... ....... ... ... ... ..... ......... ....... ...... .. 6-9 6.4 References .. ....... ...... ............ ........... ..... ................ ....... .......................... ............ 6-10 7.0 FIRE SAFETY ... ............. ..... ... .......................... ........................... .. .... ... ..... ... .......... ...... . 7-1 7.1 Fire Safety Management Measures ..... .... ....... ... .... ....... .. ........ ...... ............. .... ..... 7-2 7.1.1 Fire Prevention ....... ........ .... ....... .... ....... ... ..... ..... ... ... ........ ..... ............... 7~

7.1.2 Inspection, Testing, and Maintenance .......... ........... ... ... .... .... ..... ....... .... 7-5 7.1.3 Emergency Response Organization Qualifications, Drills, and Training ... ..................... .... .. .............. .. ... ... ..... ..... ..... ........ ..... ........ .......... 7-6 7.1 .4 Pre-Fire Planning ... ........ .... ........... ..... .... ....... ... ............................. ..... .... 7-6 7.2 Fire Hazards Analysis ..... ...... ........ ......... ...... ... ..... ............. ..... ... .. .............. ............ 7-8 VI

License Application for the American Centrifuge Plant Proposed Change 2021 7.2.1 Fire Hazards Analysis Approach .......... ........ ... ......................... ............. 7-9 7.2.2 Integrated Safety Analysis ......... ................... ..................... .... ..... ....... .. 7-10 7.2.3 Building Surveys ....................................................................... ....... 7-H lQ 7.3 Building/Facility Design ............ .................... ..... ... ............ .. ....... ......................... 7-11 7.3.1 Fire Suppression Systems ... ..... ....... ... ........ ..... ........ ....... ... ....... .... .... 7_g 12 7.3.2 Fire Alarms .................................................................................. ........ 7-13 7.4 Process Fire Safety .... .... ........ ... ... .............. ....... ....... ....... ............... ....... ......... .... ... 7-13 7.5 Fire Protection and Emergency Response ............... ........... .......... ...................... 7-14 7.5.1 Fire Protection Engineering ....... .............. ... ...... ........ ........ .. ....... ... ....... 7-14 7.5.2 Alarm and Fixed Fire Suppression Systems ................................... 74§-14 7.5.3 Firewater Distribution System .............. ... ... ...................... .... .......... . 7-1-612.

7.5.4 Mobile and Portable Equipment .......................................................... 7-16 7.5.5 Emergency Response ..... ............. ................. ...... ...................... ............ 7-16 7.5.6 Control of Combustible Materials ...................... .......... ....... ....... ....... .. 7-16 7.5.7 Use of Noncombustible Materials ..... ....... ....... ... ... ....... ........ ... ... .. .. . 7--l-+16 7.5.8 Control of Combustible Mixtures ...... ...................................... ........ 7--l-+16 7.5.9 Placement of Equipment and Operations .............................. ....... ........ 7-17 7.6 References ....... ... ....... ............................... ....... ............ .. ..... ...... ..... ....... ....... 7--l-&17 8.0 EMERGENCY MANAGEMENT ................................. ..... ..... ..... ...... ............. ....... ..... .. 8-1 8.1 High Assay Low Enriched Uranium Demonstration ............................................... 8-1 8.1.1 Nuclear Criticality ...................... ............................................................... 8-2 8.2 References .... .. ....................... .......... ........... ................................... ........................... 8-3 9.0 ENVIRONMENTAL PROTECTION .... ..... ....................................................... ........... 9- 1 Vil

License Application for the American Centrifuge Plant Proposed Change 2021 9.1 Environmental Report ..... .. ..... .... .... ............ ....... .......... ........... .............. ..... ........ 9-1 9.2 Environmental Protection Measures ... .. .... .................... ................... .... .. ..... ... 9-+-2.

9.2.1 Radiation Protection Program ... .. ....... ..... ............ .... ..... ........ .... ..... ....... 9-+.2_

9.2.2 Effluent and Environmental Monitoring .......... ... ..... ..... ..... ... ...... ..... .. 9-910 9.2.3 Integrated Safety Analysis Summary ..... ....... ....................... ....... .. ... 9-+/-425 9.3 Reports to the Nuclear Regulatory Commission .. ...... .... .... ........ ..... ..... .. ..... 9-+/-425 9.3.1 10 Code of Federal Regulations 70.59 Reports .... ....... .................... 9-+/-425 9.3.2 National Emission Standards for Hazardous Air Pollutants Reports9-+/-425 9.3.3 Baseline Effluent Quantity Reports ............................... ... ....... .... .... 9~ 26 9.4 References ....... ............ ... ...... ...................... ... ......... .............................. .... .... 9~ 26 10.0 DECOMMISSIONING ... ..... ... ..... ... ....................... ..... ............. .............. ..... ....... ......... . 10-1 10.1 High Assay, Low-Enriched Uranium (HALEU) Demonstration Program ......... 10-1

-W:+ 10.2 American Centrifuge Plant (ACP) Decommissioning Program. .......... 10-2 10.-l-.2_.1 Decommissioning Design Features ........ .... ....... ....... .......... ...... ..... .. 10-J1_

_ __ 10.2.2 Decommissioning Steps .... .. ... ...... .. ... ...... ...... ... ........ ............. ...... .. ... .. 10 1 10.2.1 Overviev, .............................................................................................. 10 5 10.2.2 Purging ................................................................................................ 10 6 10.2.3 Dismantling and Removal.. .................................................................. 10 6 10.2.4 Deeontrunina-tion .................................................................................. 10 7 10.2.5 Salvage and Sale .................................................................................. 10 7 10.2.6 Disposal ............................................................................................... 10 7 10.2.7 Final Radiation Swvey ........................................................................ 10 8 10.2.3 Management/Organization ........ ......... ......... ... ..... ........................ ... 10-+012 VIII

License Application for the American Centrifuge Plant Proposed Change 2021 10.2.4 Health and Safety .. .. .. ............................. ........... ... ....... .... ... ... ... .. .... 10-W12 10.2. 5 Waste Management .... ............................ ........ ...... .... .... ......... .. ... .. .. 10-W12 10.2.6* Security and Nuclear Material Control.. ...... .. ....... ........................ . 10-Wll 10.2.7 Record Keeping ............. ........ ........ .......... .......... ........................... . 10---l--l-_Ll_

10.2. 8 Decontamination .......... ............ ...... ................ .............. ........ ..... ... .. 10---1-2-_Ll_

10.8.1 Decontamination Service Area 10 12

10. 8.2 Procedure 10 13 10.8.3 Results 10 14 10.2.9 Agreements with Outside Organizations ........ ....... ..... ..... .... ...... .. .. 10--+416 10.2. 10 Arrangements for Funding ............... ... .................. ....... ..... ....... .... .. 10--1-16.

IX

license Application for the American Centrifuge Plant Proposed Change 2021 10 .10 .1 Plant Decommissioning Costs ......................................................... 10 14 10.10.2 UF6 Ta:ils Disposition Costs ..................... .................. ...................... 10 18 10.10.3 Total Decommissioning Liability .................................................... 10 18 10.10.4 Funding Arrangements ................................. ................ ...... ... ........... 10 19 10.-1-1--1-_References ......... ... .... .. ..... ......................................................... ..... ......... 10-2021 11.0 MANAGEMENT MEASURES ...... ... .................. ..... .... .... .. .... .. ..... ............ ........ ......... 11-1 11.1 Configuration Management .... ..... ................................... .... ...................... ...... . 11-1 11.1.1 Configuration Management Policy ... ..... ............... .. ... ... .... .... ........ ..... 11 2.

11.1.2 Design Requirements ...... ....... ..... ........ ....... ..... .. ........ .............. ........... 11 -61 11.1.3 Document Control.. ..................... .... .... .... ....... .............................. ...... 11 -1~

11.1.4 Change Control ..................... .......... ......................... ......... ... ................ 11-9 11 .1.5 Assessments .... ... ........ ........ ........ ......... ........ ........ .. ...... .. .. .... ........... .... 11-11 11.1.6 Design Verification ..... ............ ...... ........ ..... ....... ....... ... .... ... .. .... ... ... 11 --1+12 11.2 Maintenance ....................... ............ .......................... ....... .. ... ... ...... ...... ..... ... ... 11-12 11.2.1 Maintenance Organization and Administration .. ..... ......... ...... ..... ...... 11-12 11.2.2 Personnel Qualification and Training ..... ........................ ............... 11--8-14 11.2.3 Design/Work Control.. ....... ................................. ............ ................... 11-14 11.2.4 Corrective Maintenance ..... ............ ............ ..................... ........ ....... 11 ~ 16 11.2.5 Preventive Maintenance ..... .... ...................... .... .... ........... ................. .. 11-16 11.2.6 Surveillance/Monitoring ........................ ..... ........ ..... ......... ..... ... ......... 11 -16 11 .2.7 Functional Testing .... .... .......... .................................... .. ................. .... 11 -17 11 .2.8 Control of Measuring and Test Equipment.. ...... ..... ... ... .. .. ......... ........ 11-17 11.2.9 Equipment/Work History ................ ...... ..... ..... ..... ....... ... ... ... ..... ......... 11-18 X

License Application for the American Centrifuge Plant Proposed Change 2021 11.3 Training and Qualification .................................................. ........................... 11-18 11.3 .1 Organization and Management of the Training Function ... .. ... .......... 11-18 11.3.2 Analysis and Identification of Functional Areas Requiring Training ......................................................... ... ... ... ........ .......... 11 ~ 29 11.3 .3 Position Training Requirements .................................................... 11 ~ 30 11.3.4 Development of the Basis for Training, Including Objectives .. ........ 11-30 11.3.5 Organization oflnstruction, Using Lesson Plans and Other Training Guides .... .. ..... ................. .. .... .. ..... .................. ... ..... ..... ... 11-30 11.3.6 Evaluation of Trainee Learning ........ ................ ........................... 11-31 11.3 .7 ConductOn-The-JobTraining ..................................................... 11-31 11 .3.8 Evaluation of Training Effectiveness .. ......................................... 11-31 11 .3.9 Personnel Qualification .... ................................ .............. .............. 11-32 11.3.10 Provisions for Continuing Assurance ... .................... ................... 11-32 11.3.11 References ............................ ........................................................ 11-33 11.4 Procedures .. ....... ... ....................... ....................... .................. ... ... ............. .... ... 11-33 11.4.1 Types of Procedures .............. ............................. .... .. .. ... ..................... 11-33 11.4.2 Procedure Process .............. ............... .. ... ............... ... .. .... ...... .. ............ 11-36 11.4.3 Procedure Hierarchy ......................................... ......... ........ .... .... ...... .. 11-40 11.4.4 Temporary Changes ... ....................... ............................................. 11 -4G41

11. 4. 5 Temporary Procedures .... ....................................... .. .. ..................... ... 11-41 11.4.6 Periodic Review ...................... .. ................... ... ... ............................ 11 --41-42 11.4.7 Use and Control of Procedures ................... .. ........ .. ........... ........ .... 11 --41-42 11.4.8 Records .............. ......................................... .. ................................. 11 ~ 3 11.4.9 Topics to be Covered in Procedures .. ............................................ 11 ~ 3 XI

License Application for the American Centrifuge Plant Proposed Change 2021 11.4.10 References .................................. .................... ............................... 11 ~ 6 11.5 Audits and Assessments ...................... .............. .... ...... ................................... 11-46 11 .5.1 Audits ......................................... ...... ........ .......... ... ......................... 11 -4647 11.5 .2 Assessments .................................. ..... ..... ........... ............................ 11 -464 7 11.6 Incident Investigations ................................................................... ............ 11 -4+48 11.6.1 Incident Identification, Categorization, and Notification .. ............ 11 -4+48 11.6.2 Conduct of Incident Investigations ........................ ... .... .. ............... 11 -4849 11.6.3 Follow-up Written Report .................................................... ....... ... 11 -4950 11.6.4 Corrective Actions ........ ................................................................. 11 -4950 11.7 Records Management and Document Control.. ........................ .. .... ..... ..... . 11 -49.21.

11. 7 .1 Records Management Program ...... ................. .... .. ... ...................... 11 ~ .21.

11.7.2 Document Control Program ...................... ...... ...................... ......... 11 -M55 11.7.3 Organization and Administration .......................... .. ............ .. ...... .. . 11 -1/457 11.7.4 Employee Training ................................................ ........ ....... .......... 11 --:S-+58 11.7.5 Examples of Records ..................... ............ ..... ..... .... .. .... .. .............. 11 --:S-+58 11.8 Other Quality Assurance Elements ........................................... .. .............. . 11-6-1-63 APPENDICIES APPENDIX A .............................. .............. .................. .... .......... ....... ............. ... ........................ A-1 APPENDIX B ............................................................................................................................ 8-l APPENDIX C ................ ........................... .. ... .. .... ...... ...... ................. ........ .. ....... ........................ C-1 APPENDIX D .............. ............. ............................ .................... .... ..... .. .... .. .... ...... ...... ... .......... .. D-1 APPENDIX E ..................................................................................................................... ...... . E-1 APPENDIX F .................... ....... .. ...................... ............................. ............. .. .... ... ..... ..... ... .......... F-1 XII

License Application for the American Centrifuge Plant Proposed Change 2021 LIST OF TABLES Table 1.1-1 American Centrifuge Plant Major Facilities ............................................ .... 1-£ 54 Table 1.2-1 Commercial ACP Possession Limits for NRC Regulated Materials and Substances ................................................................. D-2 Table 1.2-2 HALEU Demonstration Program Possession Limits for NRC Regulated Materials and Substances ................................................................. D-6 Table 1.2+ .1_ Commercial ACP Authorized Uses ofNRC-Regulated Materials .............. 1~ 11 Table 1.2-4 HALEU Demonstration Program Authorized uses of NRC Regulated Materials ................................................................................ 1-74 Table 1.3-1 Historic and Projected Population in the Vicinity of the DOE Reservation ................................................................................................ 1-99100 Table 1.3-2 Precipitation as a Function of Recurrence Interval and Storm Duration for the DOE Reservation ............... ...................................... ... .............. ...... 1-99100 Table 1.3-3 Comparison of Flood Elevations of the Scioto River near the DOE Reservation with the Nominal Grade Elevation ...................................... 1--1-00101 Table 1.3-4 Regional Stratigraphic and Hydrogeologic Subdivisions ........................ 1--1-00101 Table 4.6-1 Contamination Levels ........................................................................................ 4-8 Table 4. 7-1 Routine Contamination Survey Frequencies ................... ... ... ..... .. .... ............... 4-13 Table 4.7-2 Bioassay Program ............................................................................................ 4-14 Table 4.7-3 Internal Dosimetry Program Action Levels .. .. ................................................. 4-15 Table 4. 7-4 DAC and Airborne Radioactivity Posting Levels .................. .. ........................ 4-15 Table 4.8-1 Posting Criteria ............................................................ .. .... .. ..... ....................... 4-20 Table 4.8-2 Radiological Protection Instrumentation and Capabilities .............................. 4-22 Table 5.4-1 Sample of Benchmarks Groups Chosen for HALEU Demonstration ........ F-3~

Table 7.1-1 Applicable National Fire Protection Association Codes and Standards .......... 7-%1 Xlll

License Application for the American Centrifuge Plant Proposed Change 202 J Table 9.2-1 American Centrifuge Plant Action Levels for Radionuclide Effluents .. ..... 9-+/-628 Table 9.2-2 Baseline Effluent Quantities for American Centrifuge Plant Discharges .... 9~ 29 Table 9.2-3 Anticipated Gaseous Effluents ..................................................................... 9-+/-%30 Table 9.2-4 Anticipated Liquid Effluents ........................................................................ 949-11 Table 9 .2-5 Environmental Baseline Activities/Concentrations, 1998-2002 .................. 9~ 32 Table 9.2-6 Environmental Baseline Activities/Concentrations, 1998-2002 .................. 9~ 34 Table 9.2-7 Environmental Baseline Activities/Concentrations, 1998-2002 .................. 94436 Table 9.2-8 Environmental Baseline Radiation Levels, 1998-2002 .. ... ............................ 9-3.8_6 Table 9.2-9 Potentially Applicable Consents for the Construction and Operation of the American Centrifuge Plant ................ .. .......................... ......... .. ................... 9~ 39 Table 10.2.2-1 Components for Potential Decontamination/Disposal at Decommissioningl0-9-ll Table 10.2. 10-1 Plant Decommissioning Cost Estimates and Expected Duration ........... 10-U22 LIST OF FIGURES Figure 1.1-1 U.S. Department of Energy Reservation in Piketon, Ohio .. ........................ 1-:E-34 Figure 1.1-2 American Centrifuge Plant Layout.. ............... ..... .. ..... .. ............................... 14435 Figure 1.1-3 X-3001 (X-3002) Typical General Equipment and Process Flow Layout .. 1~ 36 Figure 1.1-4 Feed, Withdrawal, and Product Operations .... .. ............... .. .......................... 1-3/437 Figure 1.1-5a X-3346 Feed Equipment and Process Flow Layout .................................... 1~ 38 Figure 1.1-5b X-3346 Blending/Transfer Equipment and Process Flow ........................... 1~ 39 Figure 1.l-5c X-3346 Product Withdrawal Equipment and Process Flow ........................ 1--W40 Figure 1.1-5d X-3346 Tails Withdrawal Equipment and Process Flow ....................... .. .... 1-4G4 1 Figure 1.1-5e X-3346 Typical General Equipment and Process Flow Layout .................. 1-4+42 Figure 1.1-6 X-3346A Typical General Equipment and Process Flow Layout ............... 1-4+/-43 XIV

License Application for the American Centrifuge Plant Proposed Change 2021 Figure 1.1-7 X-3344 Typical General Equipment and Process Flow Layout ......... ... .... .. 1-4;44 Figure 1.1-8 X-7725 Typical General Equipment and Process Flow Layout .. ....... ....... .. 1-4445 Figure 1.1-9 X-7727H Typical General Equipment and Process Flow Layout .. ... .. .. .... .. 1-#46 Figure 1.1-10 X-2232C Typical General Equipment and Process Flow Layout ....... ....... .. 1-4647 Figure 1.1-11 Separation Element ...... .... .. ........ ... .. .. ......... ....... ................ .... ... ....... ..... ........ 1-4+48 Figure 1.1-12 Centrifuge Schematic .... ............. ....... .... ....... .... .... ..... ............ .......... .... ......... 1-4&49 Figure 1.1-13 Example Cascade and Stage Flow Schematic .. ......... ....... ....... ..... .... ... .... ... .. 1-4950 Figure 1.1-14 Systems Interface .. ... .... .. ... ......... ...... .. .......... .. ........ ...... ... .. .... ... .......... ........ .. 1-W_ll Figure 1.1-15 Purge and Evacuation Vacuum System Schematic ....... .... ... ..... ......... ... ...... 141-52 Figure 1.1-16 Machine Cooling Water System Flow Schematic ....... ....... ..... ....... ..... ........ 1..£-53 Figure 1.3-1 Topographic Map of the Department of Energy Site-Reservation ......... 1-+0+ 102 Figure 1.3-2 Populatiofi Area Within Five-Mile Radius of the U.S. Department of Energy Reservation ...... ..... ....... ....... .... ...... ......... ... ........ ............ .. .... ... .... .. ....... ... .. 1~ 103 Figure 1.3-3 Special Population Centers Within Five Miles of the U.S. Department of Energy Reservation .. ... .... ....... .... ... ..... ......... .......................... .... .......... . 1---1-@ 104 Figure 1.3-4 Comparison of Wind Roses at 10-m Level at the U.S Department of Energy Reservation from 1998 - 2002 ..... ... .......... ...... ....... ....... ......... . 1-+G4105 Figure 1.3-5 Comparison of Wind Roses at 30-m Level at the U.S. Department of Energy Reservation from 1998 - 2002 .. .. ... ........ .. ..... .. ...... ..... ... .... ..... 1~ 106 Figure 1.3-6 Comparison of Wind Roses at 60-m Level at the U.S. Department of Energy Reservation from 1998 - 2002 .......... .............. ....... ..... ... ....... .. 1--146107 Figure 1.3-7 Location of Rivers and Creeks in the Vicinity of the U.S. Department of Energy Reservation ........ .... .. .. ...... ....... ......... ........ ...... .... ... .... ... .......... .. 1---141 108 Figure 1.3-8 Ponds and Lagoons on the U.S. Department of Energy Reservation ...... 1--l-OS 109 Figure 1.3-9 Elevations of Roadways and of the Surrounding Areas of Main Process Buildings ........ ... ...... ......... ....... .... ...... ............. ...... ..... ..... ....... .... .. 1--W911 0 xv

License Application for the American Centrifuge Plant Proposed Change 2021 Figure 1.3-10 The 10,000-year Intensity Versus Duration Graph for Storms at U.S. Department of Energy Reservation .......... .... .. .... ...... ... ......... ......... ... .... ........................ 1---l--1-0ill Figure 1.3-11 Location of the Ancient Newark (Modem Scioto) and Teays Valleys in the U.S. Department of Energy Reservation Vicinity ....... ...... ....... ..... 1-++-1-112 Figure 1.3-12 Geologic Cross Section in the U.S. Department of Energy Reservation Vicinity ................................................................................ ..... .... ..... ..... . 1-+H 113 Figure 1.3-13 Geologic Column at U.S. Department of Energy Reservation ... ............ 1--1-- 114 Figure 2.1-1 American Centrifuge Organization Chart ............................ ... ...... .... .... ... .... 2-1-+/-U Figure 9.2-1 Locations of American Centrifuge Plant Monitored Vents ............ .. ........... 9..!j!j.57 Figure 9.2-2 Locations of American Centrifuge Plant Outfalls Discharging to Waters of the United States ................ .......... ... ... ....... ................. ...... .. ...................... 9--S658 Figure 9.2-3 Locations of Soil and Vegetation Sampling Points .. .... ....... .. .... .......... ... .... . 9~ 59 Figure 9.2-4 Locations of Surface Water Sampling Points ....... ............ ... .. ................. ..... 9~ 60 Figure 9.2-5 Locations of Stream Sediment Sampling Points .............. .... ... ......... ..... .. ..... 9-S-961 Figure 9.2-6 Locations of Environmental Thermoluminescence Dosimeters on the U.S.

Department of Energy Reservation ...... .... ...................... .......................... .... 9-6G62 Figure 9 .2-7 Locations of Environmental Thermoluminescence Dosimeters Outside the U.S. Department of Energy Reservation Boundary ......... .......... ... ... .... .. 9-6+63 Figure 10 .2. 1-1 Commercial ACP Contamination Control Zone ........................................... 10-4 XVI

License Application for the American Centrifuge Plant Proposed Change 202 1 Blank Page XVII

License Application f or the American Centrifuge Plant Proposed Change 2021 ACRONYMS AND ABBREVIATIONS ACE American Centrifuge Enrichment, LLC ACH American Centrifuge Holdings, LLC ACL Administrative Control Level ACM American Centrifuge Manufacturing, LLC ACO American Centrifuge Operating ACP American Centrifuge Plant ACR Area Control Room ACS Access Control System ACT American Centrifuge Technology, LLC AEA Atomic Energy Act AHJ Authority Having Jurisdiction AIHA American industrial Hygiene Association ALARA as low as reasonably achievable amsl above mean sea level ANS American Nuclear Society ANSI American National Standards Institute ARA Airborne Radioactivity Area ARF airborne release fraction ASME American Society of Mechanical Engineers AST above ground storage tank ASTM American Society for Testing and Materials BCS Boundary Control Station BDC Baseline Design Criteria BEQ Baseline Effluent Quantity BOP Balance of Plant BUSTR Bureau of Underground Storage Tanks CA Contamination Area CAA Controlled Access Area CAAS Criticality Accident Alarm System ccz Contamination Control Zone CEDE Committed Effective Dose Equivalent GER Compliance Evaluation Reports CERCLA Comprehensive Environmental Response, Compensation, and Liabilities Act CFR Code of Federal Regulations CM Configuration Management CVP Cylinder Valve Protectors cw chilled water CWA Clean Water Act CWIP Construction Work in Progress D&D decontamination and decommissioning DA Design Authority DAC Derived Air Concentration DBE design basis earthquake DCP double contingency principle XVIII

License Application for the American Centrifuge Plant Proposed Change 2021 DFP Decommissioning Funding Plan D-G diesel generator DID defense in depth DOE U.S. Department of Energy DOT U.S. Department of Transportation DP Decommissioning Plan DR damage ratio DSA Decontamination Service Area DUF6 depleted uranium hexafluoride ECS Engineering Consulting Services EOC Emergency Operations Center EPA Environmental Protection Agency EPC Engineering, Procurement, and Construction EPCRA Emergency Planning and Community Right to Know Act ERPG Emergency Response Planning Guidelines ER Environmental Report EV evacuation vacuum FBP Fluor-BWXT Portsmouth FCA Fixed Contamination Area FHA Fire Hazards Analysis FM Factory Mutual FNAD Fixed Nuclear Accident Dosimeters FNMCP Fundamental Nuclear Material Control Plan FPPA Farm Protection Policy Act FSU former Soviet Union FWLA Fugro, Williams, Lettis and Associates FHA Fire Hazards Analysis FNAD Fixed Nuclear Accident Dosimeters FNMCP Fundamental Nuclear Material Control Plan GCEP Gas Centrifuge Enrichment Plant GDP gaseous diffusion plant GET General Employee Training GTC Gas Test Stand Center HA Hazard Analysis HALED High Assay Low Enriched Uranium HAZCOM hazardous eommunieation HAZMAT hazardous material HCA High Contamination Area HE Hazard Evaluation HEPA high efficiency particulate air HEU high enriched uranium HMTA Hazardous Materials Transportation Act HP Health Physics HRA High Radiation Area HVAC Heating, Ventilation, and Air Conditioning IC initial condition XIX

License Application for the American Centrifuge Plant Proposed Change 2021 ICP/MS Inductively Coupled Plasma/Mass Spectrometry IDS Intrusion Detection System IEEE Institute of Electrical and Electronics Engineers IEU intermediate enriched uranium IHS Industrial Hygiene and Safety IPP Interconnecting Process Piping IROFS items relied on for safety ISA Integrated Safety Analysis ISTP Integrated Systems Test Plan LCC local control center LEC Liquid Effluent Collector LEPC Local Emergency Planning Commission LEU low enriched uranium LLMW low level mixed waste LLRW low level radioactive waste LPF leak path factor LSDA Lower Suspension and Drive Assembly M&TE measuring and test equipment MAR material at risk MCNP Monte Carlo n-particle MCS Mid-America Conversion Services, LLC MCW machine cooling water MDA Minimum Detectable Activity MEI Maximally Exposed Individual MM Modified Mercalli MSDS/SDS Material Safety Data Sheet/Safety Data Sheet NA not applicable NAAOS National Ambient Air Quality Standards M&TE measuring and test equipment NCS Nuclear Criticality Safety NCSE Nuclear Criticality Safety Evaluation NDA Nondestructive Assay NEMA National Electrical Manufacturers Association NEPA National Environmental Protection Act NESHAP National Emissions Standards for Hazardous Air Pollutants NFPA National Fire Protection Association NHPA National Historic Preservation Act NIOSH National Institute for Oeeupational Health and Safety NIST National Institute of Standards and Technology NMC&A Nuclear Materials Control and Accountability l'fMMSS Nuclear Materials Management and Safeguards System NPDES National Pollutant Discharge Elimination System NPH natural phenomena hazard NRC U.S. Nuclear Regulatory Commission NSPS new source performance standards NVLAP National Voluntary Laboratory Accreditation Program xx

License Application for the American Centrifuge Plant Proposed Change 2021 OAC Ohio Administrative Code OEPA Ohio Environmental Protection Agency OJT on-the-job training ORC Ohio Revised Code OSHA Occupational Safety and Health Administration PA Public Address PBT Performance Based Training PCF Plant Control Facility PFPE polyfluoropolethers PGA peak ground acceleration PGDP Paducah Gaseous Diffusion Plant PBT Performance Based Training PHA Preliminary Hazard Analysis PM preventive maintenance PMF Probably Maximum Flood PMT post-maintenance testing PORTS Portsmouth Gaseous Diffusion Plant PPE personal protective equipment PSD prevention of significant deterioration PSM Process Safety Management PSP Protective Shipping Packages PSRC Plant Safety Review Committee PSS Plant Shift Superintendent PT performance testing PTI permits-to-install PV purge vacuum QA Quality Assurance QAPD Quality Assurance Program Description QC Quantity Control QL Quality Level ORA Quantitative Risk Analysis R&D research and development RIA Recycle/Assembly RA Radiation Areas RAM random access memory RCRA Resource Conservation and Recovery Act of 1976 RCW recirculating cooling water REIRS Radiation Exposure Information Reporting System RF respirable fraction RG Regulatory Guide RGA Regional Gnwel Aquifer RHW recirculating heating water RM river mile RMA Radioactive Material Area RMC Ridge Mast Crane RMDC Records Management and Document Control

License Application for the American Centrifuge Plant Proposed Change 2021 RMP Risk Management Program RP Radiation Protection RPM Radiation Protection Manager RQ Reportable Quantity RWP Radiation Work Permit SAR Safety Analysis Report SARA Superfund Amendments and Reauthorization Act SCBA self contained breathing apparatus SERC Ohio State Emergency Response Commission SHPO Ohio State Preservation Officer SIC standard industrial classification SME Subject Matter Expert SNM special nuclear material SPCC Spill Protection Control and Countermeasures SRD System Requirements Document SRP Standard Review Plan SSCs structures, systems, and components STP Se .vage Treatment Plant 1

SWPP Storm Water Pollution Prevention TDAG Training Development and Administrative Guide TEDE Total Effective Dose Equivalent TLDs Thermoluminescence Dosimeters TLV Threshold Limiting Value TPO threshold planning quanitity TQs Threshold Quantities TRM Training Requirement Matrices TSD Treatment, Storage, or Disposal TWC Tower Water Cooling T\VCR Tower Water Cooling Return TWCS Tower \Vater Cooling Supply UCNI Unclassified Controlled Nuclear Information UCRS upper continental recharge system UL Underwriters Laboratories UPS uninterruptible power supply USA Upper Suspension Assembly USACE United States Army Corps of Engineers USEC USEC Inc.

USGS U.S . Geological Survey USL upper safety limit UST underground storage tank VHRA Very High Radiation Area WCA workers in the controlled area WI/CL What-if/Checklist WRA workers in the restricted area XXII

License Application for the American Centrifuge Plant Proposed Change 2021 DEFINITIONS Heeling - The process for removing the residual quantity of uranium material that remains in a cylinder after routine evacuation procedures.

Natural Uranium - Any uranium-bearing material whose uranium isotopic distribution has not been altered from its natural occurring state. Natural uranium is nominally 99.283 percent 238 U, 0.711 percent 235 U, and 0.006 percent 234U (by weight relative to total uranium element).

Normal Uranium - Any uranium-bearing material having a uranium isotopic weight distribution that can be described as being (1) 0.700 to 0.724 percent in combined 233 U plus 235 U; and (2) at least 99.200 percent in 238 U.

XXIII

License Application for the American Centrifuge Plant Proposed Change 2021 Blank Page CHEMICALS AND UNITS OF MEASURE CaF2 calcium fluoride cfs cubic feet per second Ci cune cm centimeters cm2 square centimeter dpm disintegration per minute DUF6 depleted uranium hexafluoride F Fahrenheit ft feet ft/d feet per day ft2 square feet g grams Gal gallons Gal/d gallons per day HF hydrogen fluoride

m. inches keff k effective km kilometers km2 square kilometers kV kilovolts L liters lb pounds Lid liters per day lfpm linear feet per minute m meters m2 square meters mCi millicuries (one-thousandth of a curie) mCi/mL millicuries per milliliter mg milligram ( one-thousandth of a gram) mg/L milligrams per liter mph miles per hour mrem millirem (one-thousandth of a rem)

MTU metric tons uranium pCi picocurie ( one-trillionth of a curie) pCi/L piocuries per liter ppm parts per million psf pounds per square foot psi pounds per square inch REMrem roentgen Roentgen equivalent Equivalent manMan swu separative work units U30s depleted uranium oxide XXIV

License Application for the American Centrifuge Plant Proposed Change 2021 U02F2 uranyl fluoride UF6 uranium hexafluoride V volt wt. weight YA Instrument Air

µCi microcurie (one-millionth of a curie)

µCi/g m1crocur1es per gram

µg microgram ( one-millionth of a gram)

µg/kg micrograms per kilogram

µg/L micrograms per liter

µg/mL micrograms per milliliter

µg/m3 micrograms per cubic meter

µ micron or micrometer (one-millionth of a meter) 23su uranium-235 99Tc technetium XXV

license Application for the American Centrifuge Plant Proposed Change 2021 EXECUTIVE

SUMMARY

This license application was previously submitted by Centrus Energy Corp. (Centrus),

formerly known as prepared by USEC Inc. (USEC) the applicant for a license to possess and use special nuclear, source and by-product material in the American Centrifuge Plant located in Piketon, Ohio, under the Atomic Energy Act of 1954, as amended, 10 Code ofFederal Regulations (CFR) Parts 70, 40 and 30, and other applicable laws and regulations. A primary mission of the American Centrifuge technology is to provide the United States with a reliable and economical source of enriched uranium. USEC Centrus is the parent company of the United States Enrichment CorporationAmerican Centrifuge Operations, LLC (ACO). which is the current holder assignee of a sublease for portions of the Portsmouth Gaseous Diffusion Plant (GDP) reservation from the U.S. Department of Energy (DOE) through the Lease Agreement between the US. Department of Energy and United States Enrichment Corporation for the Gas Centrifuge Enrichment Plant (GCEP lease Agreement).U.S. Nuclear Regulatory Commission Certificate of Compliance for PORTS issued under 10 CFR Part 76. USEC is a global energy company and a leading supplier of enriched uranium fuel for eommereial nuclear power plants. Amerie1m Centrifuge Operating, bbGACO-(the Licensee) is a wholly owned indirect subsidiary of CentrusAmerie1m Centrifuge Holdings, LLC, ,vhieh and is a limited liability company formed under the laws of Delaware.

Amerie1m Centrifuge Holdings, LLC is a wholly ovmed subsidiary of USEC.

Deployment of the American Centrifuge Plant supports the national energy security goal of maintaining a reliable and secure domestic source of enriched uranium. Through amendments to the Atomic Energy Act, Congress created and privatized the Corporation with the intention that USEC would, among other things, conduct research and development as required, evaluate alternative technologies for uranium enrichment and help maintain a reliable and economical domestic source of enriched uranium. Centrus continues that fundamental mission through its indirect subsidiary ACO (the Licensee).

The Licensee is responsible for the design, fabrication, installation, operation, maintenance, modification and testing of the American Centrifuge Plant. The American Centrifuge Plant is a uranium enrichment facility designed to enrich, safely contain and handle uranium hexafluoride up to 10-weight percent urani um-23 5. USEC requestedA CO currently holds a license for a term of 30 years from the start of operations. The initial modular design produces approximately 3.8 million separative work units annually. This submittal continues with modular deployment of the American Centrifuge Plant and the next phase of enrichment production, which involves deployment of a cascade of 16 centrifuges to demonstration production of high-assay, low-enrichment uranium fuel for advanced reactors. The design of the American Centrifuge Plant complies with the Baseline Design Criteria specified in 10 CFR 70.64(a) and the defense-in-depth requirements contained in 10 CFR 70.64(b).

The American Centrifuge Plant is located on U.S. Department of Energy (DOE) owned land in rural Pike County, a sparsely populated area in south central Ohio. Some of these facilities are leased to the Licensee. The DOE reservation has been studied and characterized extensively by both DOE and Centrus, formerly USEC. The facilities to be utilized for the American Centrifuge Plant, which are part of the former DOE Gas Centrifuge Enrichment Plant program, were built in the early 1980s. The existing facilities will be refurbished to accommodate the

License Application for the American Centrifuge Plant Proposed Change 202 I American Centrifuge Plant. New facilities will be constructed to house withdrawal and product operations for the commercial American Centrifuge Plant. The commercial American Centrifuge Plant operation will also use other existing site-wide services such as laboratory analysis, fire protection, security, medical, waste management and environmental monitoring .

2

License Application for the American Centrifuge Plant Proposed Change 202 1 This license application follows the format and guidelines provided in NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility. The Application is written prospectively in the present tense, representing the licensed condition. The information provided reflects the design in sufficient detail to enable a reviewer to make a definitive evaluation that the American Centrifuge Plant can be constructed and operated without undue risk to the health and safety of the public, and with no significant impact to the environment.

3

License Application for the American Centrifuge Plant Proposed Change 2021 5.0 NUCLEAR CRITICALITY SAFETY The American Centrifuge Plant (ACP) possesses large quantities of enriches uranium hexafluoride (UF6) at enrichments of up to 10 t1reight ( ,vt.) percent uranium 235 (~l:B. The 1 1 commercial ACP operation is designed to enrich and safely handle up to 10 weight (wt.)% uranium-235 (2 35 U). The HALEU Demonstration Program is designed to enrich and safely handle uranium with an operational limit less than 20.0 wt. percent 235 U; however, enrichment levels up to 25 wt.%

235 U are authorized to permit for process fluctuations which can result in higher weight percent material. The maximum acceptable enrichment is identified for each operation evaluated for nuclear criticality safety (NCS). The specific authorized uses for each class of U. S. Nuclear Regulatory Commission (NRC)-regulated material are shown in Table 1.2-;!-3 (commercial ACP operation) and Table 1.2-4 (HALEU Demonstration Program) of this license application. The Licensee is required to comply with the performance requirements of 10 Code ofFederal Regulations (CFR) 70.61. 10 CFR 70.61 (d) requires that the risk of nuclear criticality accidents be limited by assuring that under normal and credible abnormal conditions, nuclear processes are subcritical, including use of an approved margin of subcriticality for safety. It also requires that preventive controls and measures must be the primary means of protection against nuclear criticality accidents. Accordingly, these requirements are implemented through this chapter slHlllllarizes the ACP Nuclear Criticality Safety (

NCS)_Program.

In accordance with the requirements contained in 10 CFR 70.62, the likelihood and risks of an inadvertent nuclear criticality we~ e evaluated in the Integrated Safety Analysis (ISA). The evaluation consider~ed moderation events, maintenance e*tolutions, maehine upset conditions, and cylinder operations accident sequences caused by process deviations or other events internal to the facility and credible external events, including natural phenomena. Criticality Events are derived and evaluated through the process of generating Nuclear Criticality Safety Evaluations (NCSEs). In the case of the commercial ACP operation, Nuclear Criticality Safety Reports (NCSRs) were generated that will be transitioned to NCSEs prior to commencement of commercial plant operations. NCSEs will be developed based on the detailed design of the commercial ACP operationThe ISA effort documented these evaluations in NCS Reports that will in tum form the bases to develop Nuclear Criticality Safety Evaluations (NCSEs) addressing the detailed design. If changes to the 1-SANCSEs or NCSRs are identified during the development of the Jl-lCSEs, the Licensee will revise the ISA, as necessary, to include any new or updated event sequence information, identify additional double contingency controls, or credit additional items relied on for safety (IROFS). The ISA concluded that includes credible nuclear criticality accident scenarios that could be identified for the ,i\,CP *vrere controlle*d through a combination of administrati1f'e and engineered controls to assure that all nuclear processes are subcritical under normal and credible abnormal conditions. Additionally, preventative controls and measures are the primary means of protection against criticality in compliance with the performance requirements of 10 CFR 70.61 (d).

The plant has established a threshold of 1 wt. percent or higher enriched 235U and 100 grams (g) or more of 235 U for determining when an evaluation for NCS considerations of planned operations must be performed. This 100 g 23 5U mass is a minimum of a factor of Ml below the minimum critical mass,. at 10 percent ~u enrichment, regardless of whether the material is optimally moderated and fully reflectednon oily, oily, or heterogeneous for a fully reflected system.

Based on this, the value is sufficiently low to use as a threshold limit. In view of this threshold, 5- 1

License Application/or the American Centrifuge Plant Proposed Change 202 1 many of the ACP NCS Program features described in this chapter may not be required to be implemented for operations below the threshold. In this regard.As described herein, the NCS Program provides the framework for a defense-in-depth philosophy to help ensure the risk of inadvertent criticality is maintained acceptably low. The NCS Program also provides the framework and resources for evaluating plant performance in establishing NCS analyses and controls for the design and operation of a uranium enrichment plant.

5.1 Management of the Nuclear Criticality Safety Program 5.1.1 Program Elements The NCS Program described in this chapter is implemented by plant procedures. The NCS procedures address plant personnel NCS responsibilities, adherence to NCSE requirements, review and approval of fissile material operations, posting and labeling requirements, response to NCSE violations, and NCS training requirements. Controls and/or barriers that are relied on to prevent inadvertent criticalities are designated as IROFS in the ISA. The NCS Program meets the Baseline Design Criteria (BDC) requirements in 10 CFR 70.64(a)(2} concerning application of the double contingency principle in determining NCS controls and IROFS in the design of new facilities and new processes.

5.1.2 Program Objectives The NCS Program meets the requirements of 10 CFR Part 70. The objectives of the program include:

Preventing an inadvertent nuclear criticality; Protecting against the occurrence of an identified accident sequence in the ISA Summary that could lead to an inadvertent nuclear criticality; Complying with the NCS performance requirements of 10 CFR 70.61; Establishing and maintaining NCS safety parameters and procedures; Establishing and maintaining NCS safety limits and NCS operating limits for IROFS; Conducting NCS evaluations to assure that under normal and credible abnormal conditions nuclear processes remain subcritical, and maintain an approved margin of subcriticality for safety; Establishing and maintaining NCS IROFS, based on current NCS evaluations; Providing training in emergency procedures in response to an inadvertent nuclear criticality; Complying with NCS BDC requirements in 10 CFR 70.64(a)(2};

5-2

License Application fo r the American Centrifuge Plant Proposed Change 202 I

Complying with the NCS ISA Summary requirements in 10 CFR 70.65(b); aoo

_*_ Complying with the NCS ISA Summary change process requirements in 10 CFR 70. 72; and~

Complying with the reporting requirements of 10 CFR 70.52 and 10 CFR 70 Appendix A.

5.2 Organization and Administration 5.2.1 Nuclear Criticality Safety Responsibilities The ACP organization and administration are described in Chapter 2.0 of this license application. The General Manager assigns responsibilities and delegates commensurate authority to ACP managers/supervisors for the implementation and oversight of the NCS requirements. The managers/supervisors ensure that sufficient resources are available for implementation of NCS requirements-:-. The Director, Nuclear Safety is responsible for implementing the ACP NCS Program. The Director, Nuclea-r Safety reports to the Senior Vice President, Field Operations and is responsible for the direct management of the NCS functions and administration of the NCS Program on a day to day basis. The management reporting structure for the ACP is depicted in Chapter 2.0 of this license application. The Director, Nuclear Safety has direct access to the General Manager for nuclear safety matters and reports directly to the Senior Vice President, Field Operations.

The ACP organization managers are responsible for ensuring that operations involving uranium enriched to 1 wt. percent or higher 235 U and 100 g or more of 235U (hereafter referred to as fissile material operations) are identified and evaluated for NCS considerations prior to initiation of the operation. The organization managers or their designees are also responsible for ensuring NCS evaluations are requested, and for ensuring implementation of the requirements contained in the evaluations for these same operations. For those fissile material operations performed by personnel from multiple organizations, the General Manager assigns responsibility for that operation to a single organization manager or designee.

Management is responsible, in their respective operations, for ensuring that personnel are made aware of the requirements and limitations established by approved NCSEs either through pre-job briefings, required reading, training, and/or procedures (based on the complexity of the change).

These managers/supervisors are responsible for ensuring fissile material operations that do not have approved NCSEs will not be performed until the necessary approvals have been obtained.

Management is responsible for ensuring that only personnel who have received and passed NCS training as specified in ACP NCS procedures will handle fissile material.

Managers/supervisors who are responsible for one or more fissile material operations are trained in NCS and ensure appropriate personnel receive NCS training as specified in ACP NCS procedures. This training provides personnel with the knowledge necessary to fulfill their NCS responsibilities. Section 11.3. 1.4 of this license application discusses the NCS training program for those who manage, work in, or work near facilities where the potential exists for a criticality accident to occur (i.e. where fissile material handling/operations are performed).

5-3

License Application for the American Centrifuge Plant Proposed Change 2021 The fissile material operators are responsible for conducting operations in a safe manner in compliance with procedures or v,ork instructions and are required to stop operations if unsafe conditions exist.

The Director, Nuclear Safety has, as a minimum, a bachelor' s degree in engineering, mathematics or related science or equivalent technical experience, and six years nuclear experience, including six months at a uranium processing plant where nuclear criticality safety was practiced.

The Director, Nuclear Safety or designee is responsible for the administration of the NCS Program.

This includes reviewing the overall effectiveness of the NCS Program, ensuring that NCS staff members are placed, trained, and qualified in accordance with written procedures, and that NCSEs are prepared and technically reviewed by qualified NCS engineers. The NCS organization is independent of organizations that require NCSEs.

Qualified NCS Engineers and Senior NCS Engineers are responsible for performing the following functions:

Providing NCSEs for fissile material operations;

Performing walk-throughs of facilities which handle fissile material and advising appropriate management of any NCS concerns;

Participating in investigation of incidents involving NCS and in the determination of recommendations for eliminating such incidents;

Assisting in emergency preparedness planning;

Providing support to the Plant Safety Review Committee (PSRC);

Participating in the review of procedures that involve fissile material operations to ensure NCSE commitments have been effectively incorporated into operating procedures; and

Participating in the review of work packages that involve fissile material operations-te ensUfe NCSE commitments have been effectively incorporated into work package instructions. For .vork packages that are used repeatedly for the same kind ofjob, the 1

re*,iew is only necessary once. For work packages that have the NCSE commitments incorporated into an approved proeedUfe, additi011al NC8 review is not necessary, as requested.

NCS group personnel have the authority to halt any unsafe activity.

The responsibilities of Senior NCS Engineers performing technical reviews of NCSEs are specified in the NCS evaluation and approval procedure. These responsibilities include:

Verifying that sufficient information is documented to allow independent analysis by a reviewer with knowledge of the process and the NCS Program;

Verifying that credible process upsets related to criticality safety are properly identified 5-4

License Application for the American Centrifuge Plant Proposed Change 2021 and evaluated;

Verifying compliance with the double contingency principle;

Checking for accuracy; and

Verifying applicability of the calculational methods.

5.2.2 Nuclear Criticality Safety Staff Qualifications ffhe NCS Program includes training and qualifications for NCS Engineers and Senior NCS ngineers which is based on the industry best practices provided in American National Standards nstitute (ANSI)/American Nuclear Society (ANS) ANSI/ANS-8.26-2007, Criticality Saferu rt:ngineer Training and Qualification Program.

The minimum requirements for a qualified NCS Engineer are:

Bachelor' s degree in engineering, mathematics, or related science;

Familiarization with NCS by having a minimum of one year experience at aH enriched uranium processing faeilityfacility that process fissionable material where nuclear criticality safety was practiced;

Completion ofNCS-related training course and KENO V .a training course or equivalent;

Performance of at least four evaluations under the direction of a Senior NCS Engineer; and

Performance of walk-through inspections under the guidance of a qualified NCS Engineer.

The Director, Nuclear Safety can modify the minimum qualified NCS Engineer qualification requirements for personnel who have worked for a minimum of three years at other facilities as an NCS Engineer.

The minimum requirements for a qualified Senior NCS Engineer are:

Completion of the minimum requirements for a qualified NCS Engineer;

Performance of the functions of a qualified NCS Engineer;

Completion of one year as a qualified NCS Engineer; and

Approval by the Director, Nuclear Safety.

The Director, Nuclear Safety may modify the minimum Senior NCS Engineer qualification requirements for personnel who have worked for a minimum of five years at other facilities as a nuclear criticality safety engineer.

5-5

License Application for the American Centrifuge Plant Proposed Change 202 I 5.3 Management Measures 5.3.1 Procedure Requirements Operations to which NCS pertains are governed by written procedures or w:ork packages.

These procedures or work packages contain the appropriate NCS controls for processing, storing, and handling fissile material. The NCSE requirements that specify employee actions are incorporated into procedures_or 'Nork packages as work instructions and are identified. Identifying these requirements ensures changes to these requirements are not made without review and approval by NCS. The NCSE requirements are incorporated into the appropriate procedures or work packages as required by the NCS Program procedure.

New and modified procedures or w:ork packages are reviewed by the appropriate safety organizations, including NCS, as specified in the procedure for procedure control and/or 1tvork control process. NCS reviews the procedures and/or .York instructions to verify that the appropriate 1

NCSE requirements have been incorporated and to verify that the proposed operation complies with NCS Program requirements. Section 11.4 ofthis license application provides more details related to the procedure development and change process.

5.3.2 Posting and Labeling Requirements Administrative NCS limits and controls for areas, equipment, and containers are presented through the use of postings and labels as specified in approved NCSEs and procedures. Postings and labels are proposed, reviewed, and approved during the NCSE implementation process. review: and approval process. Postings and/or labels are not required for engineered controls and may not be required for administrative controls vmen those limits and controls are included in "in hand" operating procedures. These limits and controls are posted on the NCS requirements signs that are controlled and maintained according to as required by the plant NCS procedures. Approved NCSEs specify the wording for the postings. Labels are also prepared in accordance with the plant NCS procedures and used as required bydctermined during NCSEs implementation. Limits and controls are printed or written in an appropriate size, and the postings and labels are placed in conspicuous locations such that they are legible to the operator at the work location, on the specific component, item, or piece of equipment, or posted at the entrance to an operating area or storage area. The specific locations may be specified in the applicable NCSE or determined by the supervision responsible for the material.

5.3.3 Change Control A configuration management (CM) program ensures that any change from an approved baseline configuration is managed so as to preclude inadvertent degradation of safety or safeguards.

The CM Program, described in Section 11.1 of this license application, includes organization and administrative processes to ensure accurate, current design documentation that matches the plant's physical configuration. NCS controls that are IROFS are controlled as QL-2 items and NCS controls that are not IROFS are controlled as QL-3 items. The methodology for designating NCS engineered and administrative controls as IROFS is described in Section 3.1.2.3.2.7 for commercial ACP 5-6

License Application for the American Centrifuge Plant Proposed Change 2021 operations and Section 3 .1 .2.3 .2.8 for HALEU Demonstration. The CM program applies to NCS and a change control process is utilized that helps ensure that the requirements of 10 CFR 70. 72 are met, including the ISA Summary update requirements contained in 10 CFR 70.72(d)(3).

Functional and physical characteristics of opera-tionsNCS engineered controls eon-trolled for NG8 are described in NCSEs and the ISA. When an NCS engineered control isthose characteristics are required to maintain classified as an IROFS , the management measures described in the CM program associated with the QL-2 classification are applied. Some NCS controls associated with the commercial ACP operations are not IROFS and are classified as OL-3 items. When those fimetional and physical characteristics are required to maintain double contingency, but are not IROFS, the management measUres in the CM program associated with the QL 3 classification are applied. Non IROFS double contingency controls will be handled as QL 3 items.

QL-3 is a quality grouping for structures, systems, and components required to fulfill the functions and meet the requirements established by the license application. For NCS controls that rely on certain structures, systems, or components, the portions of the CM program within the QL-3 classification as described in this section, as well as the following minimum features, are applied to those structures, systems, and components:

Components are identified and controlled;

Modifications are documented and reviewed;

Change control process is applicable; Setpoints and tolerances are established for applicable components;

Engineering drawings or specifications are provided;

Procurement controls are provided; and

__ Receipt inspection is used when specified.

Components and features that are identified in the NCSEs or the ISA are analyzed to determine the "boundary" of the system, encompassing those interconnecting and/or supporting items that are essential to ensure availability and reliability . The boundaries are identified on system drawings and/or other design outputs, and the configuration is verified to be as-built. These components and features are maintained in a design control document for the building or process.

Each time a change is planned, the document is reviewed by the individual (e.g. , design authority, system engineer, operations manager, maintenance, etc.) planning the change to determine if the change affects an IROFS/NCS or double contingency control. Changes that could establish new fissile material operations or affect established fissile material operations are reviewed by NCS. The NCS Program establishes and maintains NCS safety limits and NC8 operating limits for IROFS/NCS and double contingency controls in nuclear processes and maintains adequate management measures to ensure the availability and reliability of the IROFS/NCS and the double contingency controls. Operating limits may be established during flow down ofNCS safety limits to 5-7

License Application for the American Centrifuge Plant Proposed Change 2021 ensure their continued reliability and availability.

The change control process specifies the organizations required to perform reviews of changes. Changes that affect existing fissile material operations are evaluated by NCS Ifan item is relied on for the criticality safety of an operation (i.e., is an IROF8 or a double contingency control),

it will be identified and NC8 reviews the NC8E for the specific operation and to determines if the change affects the analysis performed and the conclusions made in the NCSE. The change request will be approved by NCS only if the change does not adversely impact NCS, or once a revised NCSE has determined that the change is acceptable and meets NCS Program requirements. If a change affects the ISA Summary, it is updated appropriately. In this way, modifications to controlled operations are evaluated and approved prior to implementation and placing the affected structures, systems, or components in service.

Records management and document control (RMDC) is another element of CM and is described in Section 11. 7 of this license application. Procedures, documents, and records control programs provide for centralized control and issuance of documents essential to the maintenance of the design history, and a repository for records to verify this maintenance. NCSEs are specifically included in the index of documents that are required to be controlled.

5.3.4 Operation Surveillance and Assessment To ensure that the NCS Program is properly established and implemented, walk-throughs, assessments, and audits are utilized. These activities are specified in ACP procedures.

Operating 8NMfissile material process areas are reviewed on a regular basis through a combination of walk-throughs and reviews by work crew supervision. NCS walk-throughs of facilities that may contain fissile material operations are performed by NCS personnel to determine the adequacy of implementation of NCS requirements and to verify that conditions have not been altered to adversely affect NCS. These walk-throughs are performed as specified by the ACP NCS procedure~ on walk throughs. For example, a walk-through inspection can be performed in response to trend data, at the request of the operations personnel, or due to concerns raised by employees or NCS personnel. As a minimum, speeifie fissile material operating areas are assessed by NCS personnel via walk-through at least annually, sometimes in conjunction with the assessments discussed below. By distributing the various areas' *.valk throughs o¥er a year's time, NC8 personnel are performing a field walk through on approximately a monthly basis.

Work crew supervision provides real-time assessments of fissile material operations within their operating area to ensure NCS requirements are being adequately implemented and operating conditions have not been altered to adversely affect NCS. Fissile material operations management also performs an annual self-assessment to ensure NCS program requirements are being met in the field.

In addition to the annual self-assessments, independent internal audits of the NCS Program are conducted or coordinated by the Piketon Quality Assurance Manager as described in Section 11 .5 of this license application. The purpose of these audits is to determine the adequacy of the overall NCS Program. This includes the adequacy of the NCSEs, internal assessment programs, and 5-8

License Application for the American Centrifuge Plant Proposed Change 2021 implementation of the NCS requirements.

The results of these walk-throughs, assessments, and audits are documented and reported to appropriate management. If a condition is identified that is non-compliant with NCS program requirements, field personnel are to report the condition as directed by plant procedures. If the condition is not covered by an existing procedure, consultation with a qualified NCS engineer is required before taking any corrective action. Immediate corrective actions may be provided by the responding NCS engineer verbally or in writing. :NCS emergency response is discussed in Section 5.4.2 below and is described in more detail in Chapter 8.0 of this license application.

Managers in charge of fissile material operations are provided additional training on NCS and response to NCS deficiencies as described in Section 11.3 .1.4 of this license application and the ACP NCS procedures. Each NCS non-compliance is evaluated by an NCS engineer to determine the im act on double contin enc and 10 CFR 70.61 erformance re uirements . The "as found" field conditions are reviewed against the applicable NC8E and supporting calculational docrnnend bvailability and reliability of credited controls from the armJicable NCSE are ascertained to support the shift supervisor in determining safety significance and reporting requirements. ffhe evaluation for re ortabili of events is based on whether the controls were lost or de raded i.e. whether the were eliable or unavailable to perform their safety functions) resulting in the failure to meet 10 CFR

'70.61 erformance re uirements not based on whether the safe limit of the associated arameter NCS deficiencies are reported in accordance with the requirements contained in 10 CFR Part 70, Appendix A or other appropriate reporting requirements. Incident re orting and investigation is described in Section 11.6 of this license a lication. One-hour re ortable events are si nificant operational events that must be quickly reported. Such events typically do not require substantia valuation to determine reportability (e.g., an event involving the loss of all controls, such that a criticality accident is possible). If it cannot be determined whether an NCS incident requires eporting under paragraph (a) of 10 CFR Part 70, Appendix A, the NCS incident should be reported ithin one-hour of discove . Twen -four-hour re ortable events have less safe si ificance than one-hour re ortable events and sometimes re uire more extensive evaluation to determin eportability. The twenty-four-hour time period for reportable events is intended to allow sufficien time to make this determination. If the determination cannot be completed within this time frame,

hen the NCS incident is reoorted within twentv-four hours of discoverv. The time of discoverJ begins when a cognizant individual observes, identifies, or is notified of the NCS safety significant event or condition. A cognizant individual is an individual who, by position or experience, is bxpected to understand that the condition or event adversely im_pacts double contingency and 10 CFR 70.61 performance requirements.

The deficiency data is trended to monitor and prevent future violations. Corrective actions are taken for identified deficiencies in accordance with the Quality Assurance Program Description for the American Centrifuge Plant and the Corrective Action Program as described in Section 11.6 of this license application.,__, and rRecords of actions taken are retained in accordance with RMDC requirements described in Section 11 .7 of this license application.

5-9

License Application for the American Centrifuge Plant Proposed Change 202 I 5.4 Methodologies and Technical Practices 5.4.1 Adherence to American National Standards Institute/American Nuclear Society Standards The NCS Program has been developed to comply with the requirements of American National Standards Institute (ANSI)/American Nuclear Society (ANS) ANSI/ANS-8.1 998,2014, ANSI/ANS-8.3-1997, ANSI/ANS-8.19-2014+9%, aoo-ANSI/ANS-8.20-1991, ANSI/ANS-8.21-1995, ANSI/ANS-8 .23-2007, and-ANSI/ANS-8.24-2017, and ANSI/ANS-8.26-2007 standards as discussed in this section with the exceptions noted in Section 1.4.

5.4.2 PFoeess E';'&luatioo aod Apf)FO';'&INuclear Criticality Safety Evaluation Each operation involving uranium enriched to 1 wt. percent or higher 235 U and 100 g or more of 235 U is evaluated for NCS prior to initiation. The evaluation describes the scope of the operation, evaluates credible criticality accident contingencies, and establishes NCS requirements to maintain the operation subcritical. The evaluation process is governed by written procedures.

When an NCSE (or a change to an existing NCSE) is needed for a particular fissile material operation, a request is submitted to the NCS group to evaluate the proposed operation. Other methods for initiating an NCS change include, but are not limited to;~ 1) the engineering change process, and 2) the corrective actions process, self-assessments, and external audits and inspections.

In response to the request, an NCS evaluation may be performed or the request may be returned due to inadequate detail, the change is bounded by a current analysis, or the operation does not involve uranium enriched to 1 wt. percent or higher 235 U and with mass of 100 g or more 235 U (see Section 5.4.2.1). If necessary, a NCSE is prepared (or an existing NCSE is revised) to document the analyses performed as specified in the NCS evaluation procedure. A hazard identification process (e.g. , a "What-If' analysis) is used to identify and document potential upset conditions, or contingenc}es, presenting NCS concerns. Engineering judgment of the qualified NCS engineer may indicate the need for a more detailed study. For example, a hazards and operability study may be used if the operation is complex and involves multiple interacting systems that require substantial input from operations, maintenance, and other subject matter experts to identify the possible upset conditions. A contingency analysis is performed in which the subcriticality of a process, given the occurrence of the contingency, is assessed. This analysis demonstrates the double contingency principle for the proposed operation.

Fissile material operations must comply with the double contingency principle. The double contingency principle as stated in ANSI/ ANS-8 .1--1-998,2014, Section 4.2.2, is "Process designs should incorporate sufficient factors of safety to require at least two unlikely, inde endent and concurrent changes in rocess conditions before a criticality accident is ossible." H--fl~-tt::;-t<-t"tb-i:')

aifferent pammeters or implementing at least two controls on one parameter. The hrase "chan es

n process conditions" does not imply that reliance on two different parameters is required to satisf)1 the double contingency principle. The double contingency principle is satisfied by implementing the controls necessary to ensure at least two unlikely, independent, and concurrent ~changes in process 5- 10

License Application for the American Centrifuge Plant Proposed Change 2021 conditions would have to occur before a criticali 1s ossible. Process conditions include the characteristics of a process that have an effect on nuclear criticality safety, such as parameters,

,environment, and operations. Controls are awlied as necessary to ensure each change in process conditions is unlikely to occur.

Controls include passive engineered barriers (e.g., structures, vessels, piping, etc.); active engineered features (e.g., valves, thermocouples, flow meters, etc.); reliance on the natural or credible course of events (e.g., relying on the nature of a process to keep the density of uranyl fluoride less than a specified fraction of theoretical); and administrative controls that require performance of human actions in accordance with approved _Qrocedures or v1ork instructions, or by other means that limit arameters_within s ecified values. IH-1EWi3---onEffl-l5-i:t:reH.-ffif:Hellliffil1el--ffilc..ei:ie Application of thise double contin enc principle ensures that no single credible event can result in an accidental criticality or that the occurrence of events necessary to result in a criticality is not credible.

The NCSE will document the basis for the conclusion that a change in a process condition er Am'B:ffteff~ is "unlikely." The basis may be an engineered feature, administrative control, the natural or credible course of events, or any combination of these or other means necessary to ensure the change is unlikely to occur. Where practical, the use of explicit NCS controls will be used as the preferred approach over the reliance on natural and credible course of events. The arameters or conditions relied on and the limits must be s ecified and justified in the NCSE and controlled.

eliance on two different parameters is preferred over reliance on multiple controls on a single arameter. If rel in on two or more controls on a sin le arameter diverse i.e. different means of controlling the parameter) is preferred over redundant means of control. Management measures described in Chapter 11 .0 ofthis license application and other safety programs are sometimes used to help ensure a change in a process or parameter is "unlikely." For example, the Radiation Safety Program and/or the Fundamental Nuclear Material Control Plan may be credited with providing controls on fissile material handling; the Fire Safety Program may be credited with providing controls on combustible material loading and/or hot work activities in fissile material processing/storage areas; the Procedures Program may be credited with ensuring compliance with procedures; etc.

Where the natural or credible course of events is relied upon in whole or in part to prevent a process condition change, no specific additional controls will be necessary to maintain them. The factors that influence the process are described in sufficient detail in the NCSE as items related to NCS and prograrnmatically controlled. For items that are established, maintained, and implemented by non-NCS programs, credit for availability and reliability is established as described in Section 11.1 of this license application without the need for additional NCS controls. For situations where the NCS-credited controls do not provide adequate assurance of availability or reliability (i.e.,

situations where non-NCS programmatic and physical plant changes could adversely affect the intended criticality safety function of the items relied upon for criticality safety), specific NCS controls are established, maintained, and implemented to ensure criticality safety.

Use of the natufal and credible course of events or other means in lieu of specific administrative or engineered controls for double contingency protection requires prior NRG re:view and approval. The request for reviev1 and approYal *.vill include ajustification ohvh~' administratiYe 5- 11

License Application for the American Centrifuge Plant Proposed Change 2021 or engineered controls are not needed, a description of the proposed measures in sufficient detail to permit on lfilderstonding of their safety function, and a justification of their inherent unlikeliness.

This requirement does not apply to NCS reliance on the proper implementation of other plant programs or management measures that are described in Chapter 11.0 of this license application.

This requirement also does not apply to accident sequences determined to be non credible or those sequences v1hich do not result in a critical configuration even with the loss of both double contingency controls.

The NCS evaluation process involves a review of the proposed operation and procedures-er work instructions, discussions with the subject matter experts to determine the credible process upsets which need to be considered, development of the controls necessary to meet the double contingency principle, and identification of the assumptions and equipment (i.e., physical controls) needed to ensure criticality safety.

Engineering judgment of both the analyst and the technical reviewer is used to ascertain independence of events and their likelihood or credibility. The basis for this judgment is documented in the NCSEs. Depending on the complexity of the operation, analytical methods such as Fault Tree and Event Tree Analyses may be used in the evaluation process to examine potential accident scenarios. When needed to support the analytical method, qQualitative or quantitative estimates of event frequency are-may be developed to support the determination of the likelihood of an event.

Once the NCSE is completed, a technical review of the evaluation is performed and documented. The technical review of an NCS evaluation is performed by a Senior NCS Engineer or is-an NCS Engineer completing the technical review under the guidance of a Senior NCS Engineer.

The NCSE documents the NCS requirements for the operation. The NCS requirements include the process conditions that must be maintained to meet the double contingency principle or preserve the documented basis for criticality safety and restrict the modes of operation to those that have been analyzed in the NCSE. The requirements to be included in operating procedures and/er v1ork instructions, and postings are identified.

The NCSE approval process fffSt-involves the acceptance of the NCSE by the technical reviewer. The supervisor of the affected operation also reviews the NCSE to confirm the NCSE adequately identifies normal and credible abnormal conditions and establishes requirements that are verifiable and compatible with the planned operation. A review is then performed by ti he Director, Nuclear Safety performs a review to ensure consistency with other NCSEs and other potentially conflicting requirements or regulations. After approval by the Director, Nuclear Safety, a review is performed in accordance with 10 CFR 70.72 as described in Section 11.1.4 of this license application to determine whether prior NRC approval of the NCSE is required. PSRC approval is required for initial NCSE approval and for changes that impact the ISA Summary. After initial approval, ifNRC approval is not required and the change does not impact the ISA Summary, the NCSE is reviev,ed by the responsible organiz3ation manager. Editorial changes require only the approval of the Director, Nuclear Safety. Editorial changes are defined as changes that do not change the technical basis of the NCSE. Once approved, the NCS controls, limits, evaluation assumptions, and safety items are verified to be fully implemented in the field. The 5-12

License Application for the American Centrifuge Plant Proposed Change 202 J operationsoperating organization and NCS personnel perform this verification process. The documentation of this verification process is maintained as a quality record along with the NCSE.

Management of the operating organization is responsible for implementing, through training and procedures or work instructions, the conditions delineated in the NCSE. Operational aids such as postings, labels, boundaries for fissile material operations, and fissile material movement guidelines are providedmay be used to as specified inimplement the NCSE. The manager/supervisor ensures postings and labels are prepared and verify that they are properly installed as required by the to support implementation of the NCSE. The procedures and/or work instructions are prepared or modified to incorporate the NCSE requirements. Managers/supervisors are responsible for ensuring the employees understand the procedures and/or work instructions and understand the NCS requirements before the work begins.

Each completed NCSE is issued as a controlled document. Completed NCSEs are archived and retrievable as permanent quality records in accordance with the RMDC requirements described in Section 11.7 ofthis license application. The NCSE process provides assurance that operations will remain subcritical under both normal and credible abnormal conditions.

Emergencies arising from unforeseen circumstances can present the need for immediate action. lfNCS expertise or guidance is needed immediately to avert the potential for a criticality accident, direction will be provided orally or in writing. Such direction can include a stop work order or other appropriate instructions. Documentation will be prepared within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the emergency condition has been stabilized.

New operations must comply with the double contingency principle.

5.4.2.1 Non-Fissile Material Operations Some operations involve situations in which the uranium has an enrichment ofless than 1 wt.

percent 235 U or an inventory of less than 100 g 235 U. These operations are termed "non-fissile material operations" and are performed without the need for NCS double contingency controls. The determination of which operations are fissile versus which operations are non-fissile are made by NCS and may be contained within a NCSE or as a separate document. '}/hen the determination is outside a NC8E, the determination need not be performed by a qualified NC8 Engineer. The determination of an operation being non-fissile must include normal and credible abnormal upset conditions to ensure the enrichment and/or inventory are maintained below 1 wt. percent 235 U or below 100 g 235 U. Controls are sometimes applied to a non-fissile material operation to ensure it does not inadvertently involve fissile material. - This determination is made by an NCS engineer in collaboration with the responsible line manager.

5.4.3 Design Philosophy and Review Through the CM Program, designs of new fissile material equipment and processes must be approved by NCS before implementation. Where practical, the use of engineered controls on mass, geometry, moderation, volume, concentration, interaction, or neutron absorption will be used as the preferred approach over the use of administrative controls. Advantage will be taken of the nuclear and physical characteristics of process equipment and materials, provided control is exercised to 5-13

License Application for the American Centrifuge Plant Proposed Change 2021 maintain them if they may credibly degrade such that control of the parameter is jeopardized.

The preferred design approach includes tvt'o goals. The first is to design equipment such that NCS is independent of the amount of internal moderation or fissile concentrations, the degree of interspersed moderation betv,'een units, or the thickness of reflectors. The second is to minimi2e the possibility of accumulating fissile material in inaccessible locations and, where practical, to use favorable geometry for those inaccessible locations. Passive design controls are preferred to active design controls. The preferred design approach establishes a preferred hierarchy of controls. The use of passive engineered controls; in particular, passive engineered geometry control is the most preferred. The order of preference for NCS controls is (1) passive engineered, (2) active engineered, (3) enhanced administrative, and (4) simple administrative controls. The adherence to thi-sthe preferred design approach is determined utilized during the preparation and technical review of the NCSE performed to support the equipment design. This preferred design approach is implemented as described in NCS procedures. Deviations from the preferred design approach are justified in supporting documentation to the NCSEs.

Fissile material equipment designs and modifications are reviewed to ensure that engineered controls are used for NCS to the extent practical. Administrative limits and controls will be implemented to satisfy the double contingency principle for those cases where the preferred design approach is not practical.

5.4.4 Criticality Accident Alarm System Coverage A criticality accident alarm system (CAAS) that complies with 10 CFR 70.24 and ANSI/ANSI-8.3-1997 is provided to alert personnel if a criticality accident occurs. The system utilizes an audible and/or visual signal to alert personnel in the area to evacuate to reduce radiation exposure resulting from the incident.

The need for CAAS coverage is considered during the development process for NCS evaluations. In general, coverage is provided for fissile material operations, except the UF6 cylinder storage yards as specified in Section 1.2.5 of this license application. Other exceptions to CAAS coverage are documented in NCS evaluations and are based on a conclusion in the NCSE that a criticality accident is non-credible in the area where the fissile material operation is ongoing.

Conclusions of non-credibility require at a minimum that the inventory of 235 U in the area is less than 700 g. In addition, CAAS is not required for areas having material that is either packaged or stored in accordance with 10 CFR Part 71 or specifically exempt according to 10 CFR 71.U £ . Areas that do not contain fissile material operations do not require a NCSE and do not require CAAS coverage.

The CAAS is designed to detect gamma neutron radiation levels that would result from the minimum criticality accident of concern as defined by ANSI/ANS 8.3-1997 and to provide an audible evacuation alarmannunciation by audible evacuation alarms that are supplemented by visual alarms in some areas, such as high-noise areas . A secondary function is to activate the building radiation warning lights and alarms at the X-3012 Process Support Building Area Control Room (ACR).,_and the X 1020 Emergency Operations Center 5-14

License Application for the American Centrifuge Plant Proposed Change 202 I For each area requiring CAAS coverage, a monitoring system is installed that provides coverage of the area by at least twe-one

detection units, each with the ability to actuate the alann. This arrangement allows for one detection nnit to be temporarily out of service with fissile operations continuing nnder the coverage of the other detection unit. A detection unit is a set of at least three neutron sensitive radiation detectors that may be co-located or may be distributed over the area. The detection logic of the system requires that two of the three neutron detectors must be activated to initiate the building evacuation alarm system. Each detector may be logically part of more than one detection unit.

The building evacuation alarm system includes interior CAAS evacuation horns and exterior radiation warning lights to deter personnel from entering the building area after an evacuation. In addition, facilities within ~ 125 feet of a mt-H6-l~ fffieljjtfi'- fissile material operation area requiring CAAS coverage have radiation evacuation horns installed inside and radiation warning lights installed on the exteriorto prompt evacuation and deter persoill}el from entering the area. Personnel who have routine access to these facilities have been trained to recognize and respond to these indications as described in Section 11.3 .1 .1.2 of this license application.

To protect against the loss of coverage, the CAAS includes redundant decision logic, a backup power supply, detector status information and system self-diagnostic information are provided to the X-3012 building ACR and X 1020 building. The CAAS has been designed to survive and/or withstand credible abnormal events as described in the accident analysis for a sufficient time to warn personnel to evacuate. In the event CAAS coverage is lost for an operation, plant procedures provide for compensatory actions, which may include shutdown of equipment, limiting access, halting movement of uranium-bearing material, or other actions, such as use of personal alarming dosimeters for personnel that must access the area during a CAAS outage.

Potential criticality accident locations and predicted accident characteristics are evaluated and documented in sufficient detail to assist in emergency planning as described in ANSI/ANS-8.23-2007. Additional information provided by the CAA.8 includes a h:istorical log of events and the capability to monitor and record the criticality accident for managing the post accident situation and any remedial action. Nregarding nuclear accident planning and response is discussed in 8ectionChapter 8 2.2.4 of the Emergency Plan for the American Centrifuge Plant.

5.4.4.1 Portable CAAS In the event a fissile material operation requiring CAAS coverage is performed beyond the detection range of established CAAS instrumentation, a portable unit may be used. The portable unit has the same detection capabilities as the permanently installed units, although those capabilities may be based on gamma radiation. Alarm annunciation, however, is usually limited to the immediate area within the audible range (confirmed to 65 feet or more) of the unit's alarm with an additional telemetric link to the X-3012 ACR and X 1020. This link will transmit the location of the unit, if mobile, and allow the use of the plant PA system to warn personnel within 125~ feet of the area of the portable unit to evacuate. A portable unit will not be used for more than 24 continuous hours and it may be located indoors, outdoors, or on a vehicle.

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License Application for the American Centrifuge Plant Proposed Change 2021 If fissile material operations in an area without a permanently installed CAAS are required to exceed 24 continuous hours, all personnel not directly involved in the affected operations, or otherwise required for the safety or security of the facility, will be evacuated from an area within a M 125 foot radius of the fissile material until the operations are concluded. In addition, affected operations shall be terminated as soon as safely achievable.

5.4.5 Technical Practices 5.4.5.1 Application of Parameters Provided below are general criteria associated with apQlication of nuclear _parameters.

Each parameter is assumed to be at its optimal or most reactive credible value unless specified controls are implemented to limit the paran1eter to a particular range of values.

~ When process variables can affect the normal or most reactive credible values of parameters, controls to maintain the variables are established, and the basis for the correlation betw<;en the process variable and associated controlled parameter is documented.

When instrumentation is relied on for measuring a parameter credited for NCS, instrumentation subject to facility management measures is used.

When measurement of a single parameter is used as the sole bas_is for douJJle contingency, independent means of measurement are used .

Safety limits on controlled parameters are established and/or im_plemented with sufficient margin _to account for tolerances and t¥1Certainties.

The nuclean arameters hich can im act nuclear criticali safe are summarized below, along with examples of how the parameters are controlled at the ACP. More detail on the technical practices associated with evaluating and implementing controlled parameters is provided in the NCS program procedures.

Moderation Water is considered to be the most efficient moderator commonly found in the ACP. This is because o timall moderated U02F2/water solutions are more reactive than (Reference

~ 16). When moderation is not controlled either optimum moderation or worst credible moderation is assumed as the normal case when performing analyses. When moderation is controlled, credible abnormal process upset conditions determine the worst-case moderated conditions. Generally, moderation control is not maintained by measurement; however, when used, dual independent sampling methods are implemented.

Moderation control is applied to prevent moderators (other than moderation due to air in-leakage) from entering plant equipment containing UF6. In areas where greater than the safe mass of uranium (as defined below) is handled, processed, or stored and moderation controls are applied, that facility's pre-fire plan (reference Section 7.1.4 of this license application) includes any unique firefighting strategy or tactics that may be needed to limit the use of moderator material. However, 5-16

License Application for the American Centrifuge Plant Proposed Change 2021 even in these areas, the application of the double contingency principle ensures the worst credible loss of moderation control cannot result in a critical configuration without an additional independent and concurrent upset event.

The centrifuge process equipment is comprised of a variety of closed systems designed to process gaseous UF 6- This closed system prevents minimizes the introduction of moderation due to wet air in-leakage. Also, bBecause UF6 reacts chemically with moisture (a moderator) to produce solid uranium-bearing compounds that impedes the proper operation of the process equipment, the UF6 bearing systems are designed to minimize introduction of moisture.

Moderating materials can be present as interstitial moderators that are in solution or intermixed into the fissionable material compound (e.g., water in uranyl fluoride solution). Moderating materials may also be present as interspersed moderators that exist as moderating materials located between distinct lumps or regions of fissionable material (e.g. sprinkler activation). Interstitial Interspersed moderation issues are discussed in the Reflection section, below.

Volume Volume limits are used as specified in NCSEs. The bases for volume limits are provided in each NCSE prepared for those operations requiring containers. Specific details of these bases can be obtained by referring to the applicable NCSE. When volume control is used, the size of the containers or equipmen~ is ensured through the CM Program and/or by procedurally requiring the use of certain containers for fissile material operations.

Interaction Interaction is controlled by spacing items bearing fissile material when those items could result in a criticality accident if not properly spaced. The spacing necessary to maintain a safe array of fissile material units is determined in the NCSE performed for the array . The amount of spacing needed between items is determined based on analysis of the normal and credible abnormal process upset conditions for the particular operation. The basis for the spacing is documented in NCSEs. In accordance with the preferred design approach, described in Section 5.4.3 of this chapter, passive engineered controls are used to the extent possible to ensure spacing requirements are maintained.

When used, the structural integrity of the spacers or racks is sufficient to maintain spacing for normal and credible abnormal upset conditions.

Geometry Geometry control is applied by limiting equipment dimensions for those systems that depend on the geometry for criticality safety. The geometry is determined in the NCSE that is performed for each system and depends on the normal and credible abnormal process upsets conditions related to the specific system. Geometry controls are specified in the NCSEs, are maintained by the CM Program, and are verified prior to authorizing initial operation. "Safe geometry" is a term typically used to describe systems that are not dependent on any other nuclear parameter for criticality safety.

"Favorable geometry" is a term typically used to describe systems that rely on one or more stated parameters to maintain criticality safety. However, the use of these terms is not rigidly applied 5- 17

license Application for the American Centrifuge Plant Proposed Change 2021 throughout the available literature. Both "safe geometry" and "favorable geometry" dimensions may be obtained from established standards or operation specific reactivity calculations.

Mass controls are applied on a case-by-case basis depending on the fissile material operation involved. The acceptable mass is determined based on the specific NCSE performed for the operation. The safe mass value depends on many factors including the geometry, the 235 U enrichment, composition, etc. Safe mass values may be obtained from established standards or operation specific reactivity calculations. "Safe mass" is defined as being not more than 43 .5 percent of minimum critical (k.,w ::: 1. 0) mass for specific system conditions of enrichment, geometry, moderation, reflection, etc. the quantity of fissile material that is safely subcritical under the most reactive credible conditions (defined for a given isotopic composition and physiochemical form), including allowance for over atching. Experimental data is not used as the sole source for safe mass values. Safe mass values are chosen to ensure no single credible upset can result in a critical configuration. When safe mass values are dependent on the geometry, enrichment, composition, or some other parameter, the combination of mass and the other parameter is used as one control to meet the double contingency principle.

The safe mass values are communicated to the operating personnel via the operating procedures and,lor work paekages._Unless specifically controlled, an item containing enriched uranium is assumed to contain the most 235 U credible based on the available volume. When mass is determined through measurement, instrumentation that is subject to management measures is used.

Enrichment The maximum 235 U enrichment for each o eration is established b the s ecific NCSE. +he NC8E specifies the maximum acceptable enrichment for each operation. Credible process upse conditions that could alter the 235 U enrichment are also considered in the NCSEs. When th~

enrichment of uranium needs to be measured for an NCS control, the measurement is olJt<!-ined using either installed equipment or based on samples analyzed in a laboratory.

Uranium-containing material in the ACP with 235U enrichment less than 1 wt. percent is considered incapable of supporting a nuclear chain reaction, but interaction of such materials with materials of higher enrichment is taken into consideration in the specific NCSE for those operations which involve material enriched to greater than 1 wt. percent.

The maximum 235 U enrichment of UF 6 in the ACP HAL EU cascade is W limited to less than 20 wt. percent with the potential for momentary enrichment transients up to 25 wt. % 235 U during HALEU cascade operations. Small quantities of greater than W 20 wt. percent 235 U may also be resent outside of plant e ui ment in the form ofa-at=,e.t,EtteFV-cSaFRm~erstandards. 8ome buildings on the reservation may be used to process and,lor store fissile material from both the i\CP an~

r ortsmouth Gaseous Diffusion Plant (GDP). Although the GDP has historically processed materia, at greater than 2 10 1tvt. percent gsU, this material is no longer readily available to internet with ACP operations. Hovt'ever, for conservatism, some operations in these common buildings may be unalyz:ed at greater than 2 10 wt._J)ereent gsU enriehment._HALEU Demonstration does not involve 5-18

License Application for the American Centrifuge Plant Proposed Change 2021 The maximwn 'U enrichment for each operation is established by the specific NC8E. The NC8E specifies the maximwn acceptable enrichment for each operation. Credible process upset conditions that could alter the 'U enrichment are also considered in the NC8Es. Due to the difficulty in obtaining reliable, real time enrichment measurements that are both accurate and precise enough to use as a NC8 control, enrichment is asswned to be the mmcimwn credible for each operation. Vlhen the enrichment of uraniwn needs to be measured for ag NC8 control, the measurement is obtained using either installed equipment or based on samples analyzed in a laboratory.

Density The density of materials used in a given operation is justified in the NCSE for the operation being considered. If the density must be controlled to maintain compliance with the double contingency principle, it will be documented in the specific NCSE for the operation and it will be measured using instrumentation.

UF6 in the gaseous phase, at any credible pressures and temperatures existing in the plant equipment, is incapable of supporting a nuclear chain reaction even when intermixed with hydrogenous material (e.g., hydrogen fluoride [HF]). UF6 in the gaseous phase in plant equipment has low material density.

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License Application for the American Centrifuge Plant Proposed Change 202 I Heterogeneity Heterogeneous configurations are considered for those o erations that involve small fissile material and moderator regions. Means of causing inhomogeneity are evaluated and controlled as needed depending on their effect on subcriticality. Assumptions that can affect the physical scale off heterogeneity are based on observed physical characteristics.

Heterogeneous groupings may occur for the handling of small sample containers; hovrever, 10 v.rt. percent mu is assrnned for samples handled on a safe mass basis. Using the homogeneous safe mass of 10 wt. percent mu is also safe for heterogeneous 10 wt. percent mu because, at this enrichment, the homogeneous and heterogeneous minimum critical masses are close in value.

Concentration Concentration controls are used on a case-by-case basis. When the criticality safety of an operation depends solely on the concentration of fissile material, the medium is sampled twice, the samples are verified to be properly taken by a second individual, and the two sam les are inde endently analyzed as re uired by the s ecific NCSE for the o eration involved. The specific controls and details are documented in the NCSE for each operation that relies on concentration controls. No operations exist at the plant where concentration control is applied to an operation involving more than a safe mass ofuranirnn. A container with eoneentration controlled solution is kept normally closed. Preci itating agents including freezing, are controlled as necessary to ensure they do not inadvertently affect solubility or homogeneity or increase the concentration.

,AL tyQiea~rating limit is 5 gm~ liter regardless ofenriehment. ./'L concentration of

,+-t---,fl--1.,_mu per liter is considered subcritical at any enrichment, as reeogni2ed by AN8I/t\1't8 8.1 _

2014 . If, under all postulated conditions the concentration is ahv01* s less than 11. 6 g m.Ji~ f-il1~

Reflection Normal and credible abnormal reflection is considered when performing NCS evaluations.

The possibility of full water reflection is considered when performing analyses. Interstitial Interspersed moderation is evaluated with either full water reflection or water films with a bounding water density value to simulate sprinkler activation or precipitation combined with full density water blocks to simulate personnel. It is recognized that concrete can be a more efficient reflector than water, and its otential resence is considered. If special moderators such as deuterium, beryllium, or graphite, or iflarge amounts of hydrogen-rich materials (e.g., hydrocarbon oil or polyethylene, tc. are resent the NCS evaluation ensures the modeled reflection conditions remain boundin Reflection controls are used to limit the potential reactivity of a fissile material operation.

Neutron Absorption When neutron absorbers are used as NCS controls, the intended distributions and concentrations under both normal and credible abnormal conditions are maintained in accordance with the re uirements of the a licable NCSE and ANSI/ANS-8.21 -1995, [Jse o( Fixed Neutro n Absorbers in Nuclepr Facilities Outside Reactors. These requirements are: representative sampling of the neutron absorber, sampling at a frequency based on the environment to which the neutron 5-20

License Application for the American Centrifuge Plant Proposed Change 2021 absorber is exposed, analyzing of samples for all material attributes for which credit is taken in the NCSE, and periodic ins ections of fixed neutron absorbers to ensure ade uate distribution as specified in the NCSE. Soluble neutron absorbers are not credited by the ACP NCS Program.

An NCS evaluation can take credit for the neutron absorption properties of the materials (1) added specifically for the purpose of absorbing neutrons, and (2) of construction, provided an allowance has been made for manufacturing and dimensional tolerances, corrosion, chemical reactions, neutron spectra, and uncertainties in the neutron cross-sections.

5.4.5.2 Methods of Calculation Experimental Data Experimental data are not specific enough to allow evaluation of operations performed in the ACP. The generic nature of the experimental data does not address the variables present in the different operations. Hov,re:ver, e_Experimental data are used for validation of the computer code (e.g. , Km-.i:O V.a) used to perform the calculations needed to support the development ofNCSEs.

The experimental data used are discussed in the code validation report (Reference -l-l-J-4 15).

Handbooks and Standards Handbooks and standards (e.g., ANSI/ANS-8.1-2014) are also used in some cases when simple systems are being evaluated. Handbooks and standards used for ACP operations are nationally recognized throughout the NCS industry as high_quality analyses that have been confirmed through many years of use or based on experimental data. Most of the operations performed in the ACP are too complicated to be adequately addressed by data in a handbook/standard. When isolated operations are performed with small amounts of fissile material, referencing handbooks/standards is useful to support conclusions in the NCSE. Examples of the handbooks used include, but are not limited to, ARH-600, Criticality Handbook and LA-10860-MS, Critical Dimensions of Systems Containing 235 U, 239Pu, and 233 U. Other handbooks are held to similar criteria for excellence, industry acceptance, and quality of data to be used at the ACP without further verification calculations.

Because handbooks and standards tend to give minimum critical or maximum subcritical values, use of these values for criticality controls is not appropriate to meet the double contingency principle. Instead, these values are reduced such that subcriticality can be demonstrated under normal and credible abnormal conditions.

Hand Calculations Applicable methods for evaluating single units include Modified Two Group Diffusion Equation (i.e., Critical Equation), Buckling Conversion, and Comparative Analysis.

Modified Two Group Diffusion Equation - This method is applicable to, and most widely used for, solution systems.

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license Application for the American Centrifuge Plant Proposed Change 2021

Buckling Conversion - The method of buckling conversion or shape conversion is applicable to all materials.

Comparative Analysis - This method involves direct comparison of the system configurations to subcritical data from NCS handbooks.

Applicable methods for evaluating arrays include the Solid Angle Method and the Surface Density Method using unit shape factor.

Solid Angle Method - This method is applicable to solution systems. It is not useful if reflection is more effective than a thick water reflector located at the array boundary.

The conditions that must be satisfied in order to successfully apply the solid angle method are (1) k effective (keff) of any unreflected unit does not exceed 0.80; (2) each unit is subcritical when completely reflected by water; (3) the minimum surface-to-surface separation between units is 0.3 meters; and (4) the allowed solid angle does not exceed 6 steradians.

Surface Density Method using unit shape factor - This method can be used as an approximation for large arrays of identical units containing solutions and metals. This method determines the spacing and mass of units independent of the number of units.

An important feature of the Surface Density Method is that it is equally applicable to more irregular geometries.

When hand calculations are used, the specific methodology employed will be as described in "Nuclear Criticality Safety" by R.A. Kneif, American Nuclear 8oeiety, 1991 and s~eet to a total system reactivity of0.95 for aH credible off normal e>t'ents.based on industry-accepted methods (e.g.,

areal density, solid angle technique, etc.), subject to the limitations of those methods.

Computer Calculations For those eases v,here adequate references are not available, NCS computational analyses are performed, which involve the calculation ofkeff, may be used to determine whether the system will be subcritical under both normal and credible abnormal process conditions. Computer codes that simulate the behavior of neutrons in a process system or that solve the Boltzmann transport equation are used.

Computer calculations ofkeffprovide a method to relate analytical models of specific system configurations to experimental data derived from critical experiments. A critical experiment is defined as a system that is intentionally constructed to achieve a self-sustaining neutron chain reaction or criticality. Critical experiments that have specific, well-defined parametric values and are adequately documented are termed benchmark experiments. Computer codes are validated using experimental data from benchmark experiments that, ideally, have geometries and material compositions similar to the systems being modeled.

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License Application for the American Centrifuge Plant Proposed Change 2021 Validation of the computer code determines its calculational bias or uncertainty as well as the effective margin of subcriticality. The validation involves the modeling of benchmark critical experiments over a range of applicability. Because the kdl' value of a critical eJcperiment is essentially 1, the bias of the code is taken to be the deviation of the calculated values of kdl' from unity. Statistical analysis is employed to estimate the calculational bias, v.rhich includes the uncertainty in the bias and uncertainties due to extensions of the area of applicability, as well as the effective margin of subcriticality. Uncertainty in the bias is a measure of both the precision of the calculations and the accuracy of the experimental data. The validation of the computer code specifically defines the mmcimum acceptable kdl' used to determine subcriticality.

The margin of sub criticality used for the plant results in a kdl' upper safety limit that ensures that there is a 95 percent confidence that 99. 9 percent of future kdl' values less than this limit will be suberitieal. A minimum margin of suberitieality of 0.02 in kdl' is used to establish the acceptance criteria (i.e. , upper safety limit) for criticality calculations for abnormal conditions at 5 percent 'Y enrichment and below. Above 5 percent 'Y enrichment, a minimum margin of subcriticality of 0.05 in kdl' is used. Also, for normal ease calculations supporting processes that are not under moderation control, a minimum margin ofsuberitieality of0.05 in kdl'is used. Abnormal conditions are changes to a controlled parameter that result in a violation of the limit on that parameter. For example, in an operation that relies on maintaining spacing bet\veen fissile units, an error that results in the units being closer than the limit would represent an abnormal condition. Similarly, operations that rely on moderation control of UFe would be in an abnormal condition 1,vhen the moderation control 1,vas lost and operations that rely on control of 'U mass would be in an abnormal condition 1

.vhen the mass limit v1as violated.

The upper safety limit varies vrith the computer system, codes, cross sections, and materials used in the validation.

The calculation ofkdl'is accomplished by the use of computer codes that utili2:e Monte Carlo teeh..-iiques to determine keff of a system. Computer models representing the geometrical configuration and material compositions of the system are developed for use within the code. The development of appropriate models must account for or conservatively bound both normal and credible abnormal process conditions.

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License Application for the American Centrifuge Plant Proposed Change 2021 When NCS is based on computer code ealeulations of k.:.tr, controls and limits are established to ensure that the mrudmlHil }(,,ff' complies .vith the applicable code *validation for the type of system 1

being evaluated. For example, NCS related IROFS developed dur.ng initial license application were de*veloped using reactivity calculations performed on personal computers mrJ1ing the Microsoft

'Nindov,s XP operating system and validated as described in Reference 11. Generally, these calculations were performed \vith an upper safety limit of 0.955 up to5 percent ~u enrichment; however, specific cases may use a higher or lov,er limit based on equations from Table 14 of Reference 11. Above 5 percent ~u enrichment, a margin of subcriticality of0.05 will be applied to calculations performed using the personal computers described above with a resulting upper safety limit of 0.925. Reactivity calculations, performed after initial license application, comply with the code validation for the specific system used to perform the ealeulation.

Seeping and analysis calculations may be performed utilizing various unvalidated computer codes; however, computer calculations of l(df used as the basis for NCS evaluations are confirmed by, or performed using, configuration controlled codes and cross section libraries for which doclHilented validations are performed with at least the same degree of conservatism as that presented in Reference 11 and are in accordance with ANSI/ANS 8.1 1998. Calculations are performed using materials of construction and other parameters consistent *with the area of applicability described by the relevant validation report. The area of applicability used by Reference 11 covers enrichments from 2 percent to 30 percent mu enrichment with moderation le*,els from an wmu of 8 to 1,438 with an a>verage energy group of 151.7 to 220 using the 23 8 group ENDF/B V cross section library. Moderating materials from Reference 11 include water and paraffin and reflectors range from bare systems to reflection with .vater, steel, paraffin, polyethylene, concrete, 1

and lead. Other materials ineluded in the area of applicability from Reference 11 are stainless steel, zirconilHil, allHilinum, fluorine, and oxygen.

Extensions to the area of applicability are justified when using tech.-iiques described in NUREG/CR 6698. When materials of construction are used that are not represented in the area of applicability, the *N cS engineer has several options a>lailable to address that situation. First, the specific material can be left out of the model. Second, a different material can be substituted that is within the AOA and provides a similar (or more conservative) amount of neutron moderation, multiplication, or reflection. Third, the material ean be ineluded based on a reviev, of its neutron cross sections that conclude no significant impact ean occur from that material. Fourth, the material can be included but ',Yi.th an adjustment in its density so that any unknown effect is minimized. Fifth, the material ean be included with a reduction to the upper safety limit to aeeount for the additional uncertainty. Lastly, additional benchmark experiments can be added to the validation to specifically include the material. The NRG vrill be notified in the event an extension to the area of applicability

.vill not adequately encompass the parameters of interest for a specific calculation and a revision to Reference 11 is needed to establish a new area of applicability.

Prior to implementing changes to processes based on ealeulations requiring extension to the validated area of applicability as determined in the validation report, NRG review and approval shall be obtained. The request for NRG review and approval shall include a description of the change, the reason that such a change is needed, and the method used to extend the area of applicability.

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License Application for the American Centrifuge Plant Proposed Change 2021 The methodology used in a validation report involves statistical analysis to determine the bias and bias oocertainty for the critical experiments included in the validation. Guidance from NUREG/CR 6698, Guide for Validatimi aflVuelear Criticality Safety CakulationaU,fethodology, is used to perform the validation. The upper safety limit is computed by subtracting the absolute value of the bias, the bias ooeertainty, and the minimum margin ofsuberitieality from unity . Positive bias is not credited. The exact statistical technique used to obtain the bias and bias ooeertainty depends on the speeifie validation report. The techniques used in Reference 11 included the lower tolerance limit or the lower tolerance band for normally distributed data and a non parametric technique for non normally distributed data.

The computer codes and cross sections used in performing kdf calculations are maintained in accordance with a configuration control plan. Quarterly, or prior to use, one of the following is performed: a bit by bit comparison of the production version of the software (executable modules and data libraries) versus an archived production version; or a comparison of the output from all validation eases versus archived output of all validation eases from the original validation performed when the production version was installed to ensure no changes in the ealeulated kdf for the validation eases.

Changes to the hardv,*are or software are evaluated in aeeordanee with l OCFR 70. 72 change requirements. Some changes are expected to result in changes to the ealeulational algorithm and vfill require a new 11alidation. Such changes include revisions to the softv.'UFC used to calculate reactivity, updates to the cross section libraries, changes to the operating system kernel, changes to the central processing unit, or changes to the motherboard . Other changes are not expected to result in changes to the caleulational algorithm and will require only that the validation eases be re flHl and compared to the original results. Such changes include increasing the available RAM, changing a hard drive, graphics card, network interface eard, or other peripheral. In the Microsoft Windows environment, periodic changes to components of the operating system are common as Microsoft issues updates or patches to the platform. Also, instalJation and modification of software not used to ealeulate reactivity will be performed to support day to day business needs. These minor changes are not expected to impaet any reacti,*ity calculations, but to ensure this, a verification of the validation eases will be performed at least quarterly as described above.

The System Administrator, 0fl NCS engineer, is responsible for controlling access to the software.

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License Application for the American Centrifuge Plant Proposed Change 2021 5.5 References

1. ANSI/ ANS-8.1 --1-9-9&,2014 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors

2. ANSI/ANS-8.3-1997, Criticality Accident Alarm System

3. ANSI/ANS-8 .19-201 4-l-996, Administrative Practices for Nuclear Criticality Safety

4. ANSI/ANS-8.20-1991. American National Standard for Nuclear Criticality Safety Training L_ANSI/ANS-8.21-1995 , Use ofFixed Neutron Absorbers in Nuclear Facilities Outside Reactors

6. ANSI/ ANS-8 .23-2007, Nuclear Criticality Accident Emergency Planning and Response

7. ANSI/ ANS-8.24-2017, Validation ofNeutron Transport Methods for Nuclear Criticality Safety

~ 8. ANSI/ANS-8.26-2007, Criticality Safety Engineer Training and Qualification Program 6-;.2.,_ARH-600, Criticality Handbook, Volumes I, II, and III, Atlantic Richfield Hanford Co. Report (1968) :-Ul_LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Planl'I_

&ll,__LA-10860-MS, Criticality Dimensions of Systems Containing 235 [1, 239 Pu, and 233 U, 1986 Revision

~ .lb_NRC Regulatory Guide 3. 71 , Revision l 0, Nuclear Criticality Safety Standards for Fuels and Material Facilities, Revision 3

-l-(h~ NUREG-1513 , Integrated Safety Analysis Guidance Document

-l-h~ NUREG-1520, Standard Review Plan for the Re>piew (}fa License Applieatien for a Fuel Cycle Facilitisi§y License Applications, Revision 2

'hh~ W8M8 CRT 03 0093, United States Ertriehment Cerperatie,q (USEC) PC SCALE 4.4a Validatien (U), Revision 2, No:vernber 2005EE-3101-0013, NCS Code Validation o{SCALE 6.2.3 and Cross Section Set v7-252 for keu:Calculations, Rev. 0, December 2019

13. NUREG/CR 6698, Gllide for Validatie,q of}hwlear Criticality Safety Calelllatie,qa/

,~{ethedelegy, January 200 I

16. NC8 CAI£ 04 001, Sterage &JUF4 eil Afixtllres in Safe V0lllme Centainers, September 2005 DAC-3101-0006, Safe Mass Study for UF1 and Oil, February 2020

17. "International Handbook of Evaluated Criticality Safety Benchmark Experiments,"

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License Application for the American Centrifuge Plant Proposed Change 2021 NEAINSC/DOC (95) (03), Nuclear Energy Agency Science Committee, Organization for Economic Co-Operation and Development, July 2018 Edition.

18. Jordan, W.C., Landers, N.F., Petrie, L.M., "Validation of KENO V.a Comparison with Critical Experiments," ORNL/CSD/TM-238, Martin Marietta Energy Systems, Contract Number DE-AC05-84OR21400, December 1986.

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v t e Letter Sequence Supplement
EPID:L-2020-LLA-0085, Centrus Energy Corp. Final Environmental Assessment for High Assay Low-Enriched Uranium (HALEU) LAR (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement Administration Questions Answers 14 October 2020 Results Approval Other: ML21138A828, ML21148A147, ML21148A261, ML21148A291, ML21153A407
MONTHYEAR2021-04-27 [Table View]

ML21085A524

Text

Dear Mr. Lubinski:

References:

SUMMARY

SUMMARY

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