ML24135A293

American Centrifuge Plant; License Amendment Request Re Proposed Changes to the American Centrifuge Operating, LLCs License Application and Supporting Documents

ML24135A293
Person / Time
Site:	07007004
Issue date:	05/09/2024
From:	Karen Fitch American Centrifuge Operating
To:	John Lubinski Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML24135A314	List: ML24135A293
References
ACO 24-0035
Download: ML24135A293 (68)

Text

May 9, 2024 ACO 24-0035 CUI// SP-EXPT /SP-SRI// NOFORN Security-Related Inform ation - Withhold Under 10 CFR 2.390 Export Controlled.Inform ation ATTN: Document Control Desk John W. Lubinski, Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 American Centrifuge Plant; Docket Number 70-7004; License Number SNM-2011 License Amendment Request: Proposed Changes to the American Centrifuge Operating, LLC's License Application and Supporting Documents INFORMATION TRANSMITTED HEREWITH IS PROTECTED FROM PUBLIC DISCLOSURE AS CONFIDENTIAL COMMERCIAL OR FINANCIAL INFORMATION AND/OR TRADE SECRETS PURSUANT TO 10 CFR 2.390 AND 9.l 7(a) AND INFORMATION TRANSMITTED HEREWITH IS PROTECTED FROM DISCLOSURE PURSUANT TO 10 CFR PART 810

Dear John Lubinski,

In accordance with 10 Code a/Federal Regulations (CFR) 70.72, American Centrifuge Operating, LLC (ACO) hereby submits a license amendment request for the U.S. Nuclear Regulatory Commission's (NRC) review and approval. ACO proposes to amend the License Application and Supporting Documents to provide and clarify exemptions from the requirements of 10 CFR 70.24, Criticality Accident Requirements. These exemptions support both the American Centrifuge Plant and High Assay Low Enriched Uranium (HALEU) Demonstration cascade operations. provides a detailed description, justification, and ACO's significance determination for the proposed changes. Enclosure 2 provides the proposed changes to LA-3605-0001, License Application for the American Centrifuge Plant. Enclosure 3 provides the proposed changes to the NR-3605-0008, Emergency Plan for the American Centrifuge Plant. Enclosure 4 provides the proposed changes to LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant. Enclosure 5 provides the proposed changes to LA-3605-0003A, Addendum 1 of the Integrated Safety Analysis Summary for the American Centrifuge Plant - HALEU Demonstration.

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John W. Lubinski May 9, 2024 CUI// SP-EXPT I SP-SRI// NOFORN Security-Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information ACO 24-0035, Page 2 Proposed changes from the previously NRC-approved documents are noted with revision bars in the right-hand margin. Based upon the 10 CFR 70.32 and 70.72 evaluations, not all proposed changes described warrant the NRC's review and approval; however, are provided for completeness to assist in the review efforts.

Enclosures 4 and 5 contain Security-Related Information and ACO requests these enclosures be withheld from public disclosure pursuant to 10 CFR 2.390(d)(l). Enclosures 4 and 5 have also been determined, in accordance with the guidance provided by the U.S. Department of Energy (DOE), to contain Export Controlled Information and must be protected from disclosure per the requirements of 10 CFR Part 810. Additionally, Enclosures 4 and 5 also contain Proprietary Information and ACO requests that this enclosure be withheld from public disclosure pursuant to 10 CFR 2.390( a)( 4). An affidavit required by 10 CFR 2.390(b )(1 )(ii) is provided within Enclosure

6.

After the NRC staff has had an opportunity to review the enclosures, ACO is available to support any needed discussions to address questions or clarify issues. ACO respectfully requests NRC complete their review and approval on or before November 8, 2024, to support ACO's continued HALEU Demonstration cascade operations.

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

Enclosures:

As Stated Sincerely, Kelly L. Fitch Regulatory Manager Document/matter transmitted contains CUI// SP-EXPT I SP-SRI// NOFORN Security-Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information Proprietary Information When separated from Enclosures 4 and 5, this cover letter and Enclosures 1, 2, 3, and 6 are uncontrolled.

CUI// SP-EXPT / SP-SRI II NOFORN Security-Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information John W. Lubinski May 9, 2024 ACO 24-0035, Page 3.

cc (without Enclosur.es, unless otherwise noted):

C. Blanton, DOE Y. Paraz, NRC HQ (Enclosures)

A. Ford, DOE Idaho J. Grice, NRC Region II (Enclosures)

J. Hutson, Contract Support J. Lingard, DOE Idaho L. Pitts, NRC Region II (Enclosures)

M. Reim, DOE-NE T. Sippel, NRC HQ (Enclosures)

D. Woodyatt, NRC HQ (Enclosures)

Document/matter transmitted contains CUI// SP-EXPT / SP-SRI// NOFORN Security-Related Information - Withhold Under 10 CFR 2.390 Export Controlled Information Proprietary Information When separated from Enclosures 4 and 5, this cover letter and Enclosures 1, 2, 3, and 6 are uncontrolled.

of ACO 24-0035 Detailed Description, Justification, and Significance Determination Information Contained Within Does Not Contain Export Controlled Information Reviewing Official:

Lori Hawk, ACO Date:

05/09/2024 ACO 24-0035 Page 1 of 16 Detailed Description, Justification, and Significance Determination Detailed Description of Change American Centrifuge Operating, LLC (ACO) is requesting U.S. Nuclear Regulatory Commission (NRC) review and approval of the proposed changes for portions of the American Centrifuge Plant (ACP) License Application and Supporting Documents.

These proposed changes support amending the licensing documentation to provide and clarify exemptions from the requirements of 10 Code of Federal Regulations (CFR) 70.24, Criticality Accident Requirements. Additionally, general proposed changes are also included within this license amendment that did not warrant the NRC's prior review and approval; however, are provided for completeness to assist in the review efforts. Proposed changes are described below:

LA-3605-0001, License Application for the American Centrifuge Plant [Enclosure 2]:

Section 1.1.5.6.5, Criticality Accident Alarm System, was revised to reference Section 5.4.4 of the License Application and to reference that exemptions from the requirements of 10 CFR 70.24 are documented within Section 1.2.5 of the License Application.

Section 1.2.5, Special Exemptions or Special Authorizations, includes additional exemptions from the requirements of 10 CFR 70.24 addressing criticality monitoring as identified in Section 3.10.6 of LA-3605-0003, Section 3.10.4 of LA-3605-0003A, and prepared per Section 5.4.4 of the License Application. As such, the following proposed changes have been made:

° Clarified that the original exemption for transportation, handling, and storage of solid UF6 filled cylinders is specific to the commercial ACP operations.

0 ACO proposes an exemption from the requirements of 10 CFR 70.24 for the handling, storage, and transportation of fissile 30-series cylinders used for the High Assay Low Enriched Uranium (HALEU) Demonstration project. Note: The transportation aspect referred to under this exemption request is for on-site transportation following receipt of fissile JO-series cylinders (i.e., feed) from an NRC licensed shipper.

Off-site transportation of special nuclear material is not an authorized activity under the currently described HALEU Demonstration project.

0 ACO proposes an exemption from Criticality Accident Alarm System (CAAS) coverage when the CAAS or its associated equipment is out of service and compensatory measures that provide an equivalent risk reduction are in place.

0 ACO proposes an exemption from CAAS coverage for non-fissile material operations (NFMOs).

0 ACO proposes an exemption from CAAS coverage for areas in which a nuclear criticality safety evaluation (NCSE) has evaluated the fissile material operation and determined that a criticality accident is not credible. Conclusions of non-credibility require, at a minimum, that the inventory of 235U in the area is less than 700 grams.

ACO 24-0035 Page 2 of 16 0

ACO proposes an exemption from CAAS coverage for storage areas in which the only special nuclear material present is contained in packages defined in 10 CFR Part 71 or specifically exempt according to 10 CFR 71.15.

Sections 1.4.1, American National Standards Institute/American Nuclear Society, (ANSI/ ANS) and 1.4.2, American National Standards Institute, were revised to reflect the new section references provided within the LA-3605-0003 and LA-3605-0003A.

Section 5.4.2.1, Non-Fissile Material Operations, was revised to clarify that operations in which the uranium enrichment is less than 1 weight percent or an inventory of less than 100 grams 235U are termed "non-fissile material operations" and are performed without the need for NCS double contingency controls.

Section 5.4.4, Criticality Accident Alarm System Coverage, was revised to clarify that CAAS is required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from the 10 CFR 70.24 and refers to Section 1.2.5 of the License Application for more details related to the CAAS exemptions.

Section 8.1.1, Nuclear Criticality, was revised to clarify that CAAS is required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from the 10 CFR 70.24 and refers to Section 1.2.5 of the License Application for more details related to the CAAS exemptions.

Section 8.2, References, was revised to remove a reference that was not used within this chapter.

Chapters 1.0, 5.0, and 8.0 of the License Application also contain grammatical and formatting changes throughout.

NR-3605-0008, Emergency Plan for the American Centrifuge Plant [Enclosure 3]:

Section 2.2.4, Nuclear Criticality, was revised to clarify that CAAS is required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from the 10 CFR 70.24 and refers to Section 1.2.5 of the License Application for more details related to the CAAS exemptions.

Sections 2.2.4 and 11.0, References, were revised to delete an old Gaseous Diffusion Plant reference based upon the fact that this portion of the site is undergoing decommissioning activities, as well as correct the year reference related to ANSI/ANS-8.23.

Document also contains grammatical and formatting changes throughout.

ACO 24-0035 Page 3 of 16 LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant [Enclosure

.4}:

Section 3.10.6, Criticality Accident Alarm System, was revised to clarify that CAAS is required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from the 10 CFR 70.24 and refers to Section 1.2.5 of the License Application for more details related to the CAAS exemptions. Additionally, this section was revised to reference Section 5.4.4 of the License Application to clarify that the need for CAAS coverage is considered during the NCS evaluation process.

Section 3.10.6.1.1, Cylinder Storage Yard Criticality Accident Alarm System Exemption, was revised to clarify that the previously approved cylinder storage yard CAAS exemption is only applicable to the commercial ACP operation and that the exemption from CAAS requirements for fissile 30-series cylinders used for the HALEU Demonstration project is evaluated in LA-3605-0003A.

The following new sections were added to describe the exemptions documented in Section 1.2.5 of the License Application:

0 New Section 3.10.6.1.2, Criticality Accident Alarm System Out of Service Exemption 0

New Section 3.10.6.1.3, Non-Fissile Material Operations 0

New Section 3.10.6.1.4, Operations for Which a Criticality Accident is Not Credible 0

New Section 3.10.6.1.5, Special Nuclear Material Packaged-as Defined in JO CFR Part 71 Section 3.16, References, added ANSI/ANS-8.1-2014 as used in new Section 3.10.6.1.4.

Document also contains grammatical and formatting changes throughout, and the new sections have been added to the Table of Contents.

LA-3605-0003A, Addendum I of the Integrated Safety Analysis Summary for the American Centrifuge Plant - HALEU Demonstration [Enclosure 5]:

Section 3.10.4, Criticality Accident Alarm System, was revised to clarify that HALEU Demonstration has CAAS coverage as required by 10 CFR 70.24 unless the NRC has granted an exemption from the 10 CFR 70.24 CAAS requirements as documented in Section 1.2.5 of the License Application.

The following new sections were added to describe the exemptions documented in Section 1.2.5 of the License Application:

0 New Section 3.10.4.1, Criticality Accident Alarm System Exemption, was added which discusses the CAAS requirements of 10 CFR 70.24 and 70.17. Additionally, this section clarifies that requested exemptions specific to HALEU Demonstration are described within and requested exemptions that are explicitly applicable to only the ACO 24-0035 Page 4 of 16 commercial ACP or that apply to both are discussed within Section 3.10.6.2 of LA-3605-0003.

New Section 3.10.4.1.1, Handling, Storage, and Transportation of Fissile 30-Series Cylinders, was added to describe the exemption for operations involving handling, storage, and transportation of fissile 30-series cylinders for HALEU Demonstration.

Document also contains grammatical and formatting changes throughout, and the new Section 3.10.4.1 has been added to the Table of Contents.

No changes are being proposed to the currently described commercial ACP or HALEU centrifuge cascade design, authorized uses, or authorized possession limits. Additionally, no changes are being proposed to the American Centrifuge Lead Cascade Facility Materials License that is undergoing NRC's review and approval to terminate.

The proposed changes contained within Enclosures 2 through 5 are identified by the following method:

Blue Strili:eeut - Identifies text to be removed Red underline - Identifies text to be added Justification The following exemptions from the requirements of 10 CFR 70.24 addressing criticality monitoring are identified in Section 3.10.6 ofLA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant (Enclosure 4), and Section 3.10.4 of LA-3605-0003A, Addendum 1 of the Integrated Safety Analysis Summary for the American Centrifuge Plant -

HALEU Demonstration (Enclosure 5), and prepared in accordance with Section 5.4.4 ofLA-3605-0001, License Application for the American Centrifuge Plant (Enclosure 2).

10 CFR 70.24, Criticality Accident Requirements, requires that licensees authorized to possess special nuclear material in a quantity exceeding 700 g of contained 235U shall maintain in each area in which such licensed special nuclear material is handled, used, or stored, a monitoring system capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of two meters from the reacting material within one minute.

10 CFR 70.17 allows the Commission, upon application of any interested person or upon its own initiative, to grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The requested exemptions are authorized by law because there is no statutory provision prohibiting the grant of the exemption. The requested ACO 24-0035 Page 5 of 16 exemptions will not endanger life or property or the common defense and security and are otherwise in the public interest for the reasons discussed below.

The CAAS is a standalone system comprised of detectors and annunciators that do not interface with or affect property, defense, or security; therefore, the key requirement for evaluating the 10 CFR 70.24 exemption is that it "not endanger life." The activities that are mandated before and after a criticality event are the same, regardless of how the criticality is detected and personnel notified.

NDREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision 2, Section 5.4.3, states that "an applicant that does not meet applicable guidance in the standard review plan should describe and justify an acceptable alternative to meet the regulations." Section 5.4.3.1.1, states that Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Material Facilities, endorses ANSI/ANS-8 national standards, with some exceptions and qualifications. The NRC endorsement of these standards means that they provide methods and practices generally acceptable to the NRC staff for the prevention and mitigation of criticality accidents. However, application of a standard is not a substitute for detailed NCS analyses for specific operations. This section also states that Regulatory Guide 3.71 endorses ANSI/ ANS-8.3-1997 (reaffirmed in 2003) in full.

ANSI/ ANS-8.3-1997, Criticality Accident Alarm System, Section 4.1.1 states that "installation of an alarm system implies a nontrivial risk of criticality" and Section 4.1.3 states that "the purpose of an alarm system is to reduce the risk to personnel. Evaluation of the overall risk should recognize that hazards may result from false alarms and subsequent sudden interruptions of operations and relocation of personnel."

Each of the exemptions from 10 CFR 70.24 are discussed in more detail below to demonstrate that the change is justified and safe.

A. Handling, Storage, and Transportation of Fissile 30-Series Cylinders This proposed change requests an exemption from CAAS coverage for the handling, storage, and transportation of fissile 30-series cylinders as part of the HALED Demonstration project. NCSE-GEN-001, Nuclear Criticality Safety Evaluation for Large UF6 Cylinders [Revision 5], provides the requirements necessary to maintain criticality safety for the handling and storage of large DF 6 cylinders (3 0-series and 12-series) for HALED Demonstration.

Also, NCSE-GEN-001 provides an evaluation which demonstrates that the handling, storage, and transportation of solid DF6 filled 30-series cylinders presents a trivial risk of criticality. The NCSE states that operations involving handling, storage, and transportation of fissile 30-series cylinders shall be provided with CAAS coverage, unless NRC approved CAAS exemption criteria are met; therefore, the 30-series cylinders containing fissile DF6 will continue to be stored and handled within CAAS coverage until an NRC authorized exemption is obtained.

ACO 24-0035 Page 6 of 16 This exemption is based on the conclusion within NCSE-GEN-001 [Revision 5] that the risk of criticality for these operations is trivial and credits the robust design features, conservative evaluation assumptions, and multiple controls.

As discussed in Section 1.2.5 of the License Application, ACO proposes the following basis for the exemption for CAAS coverage not being required for the handling, storage, and transportation of fissile 30-series cylinders as part ofHALEU Demonstration because there is a trivial risk of criticality for these operations.

The handling, storage, and transportation of fissile solid UF6 filled 30-series cylinders are based on robust passive design features that are subject to management measures and administrative controls that further ensure cylinder integrity and reliability. The passive design features for 30-series cylinder operations include the robust design of the cylinders and storage array fixtures. The 30-series cylinder is fabricated of fire-resistant materials and is designed in accordance with ANSI N14.l. As such, 30-series cylinders are designed to not breach during a design basis fire and are resistant to damage from dropping, from contact with other cylinders and structures, and from corrosion. The controls, in combination with the robust design of the 30-series cylinders and the self-sealing nature of UF6 and water at the location of a credible crack or puncture, minimizes the accumulation of moderator within a fissile 30-series cylinder. The storage fixtures used in 30-series cylinder storage arrays are also made of fire-resistant materials and are designed to survive a design basis fire as, well as minimize lateral movement of the stored cylinders and elevate the cylinders above the ground/floor.

The administrative controls associated with the handling, storage, and transportation of fissile 30-series cylinders support the passive design features discussed above by requiring periodic inspections of cylinders to identify damage, breaches, corrosion, and leak-related accumulation of fissile materials; limiting UF6 enrichment in the cylinders; prohibiting stacking of cylinders; establishing spacing from other fissile materials or processes; protecting the cylinders from conditions that could cause damage; and protecting the cylinders from conditions that could lead to water intrusion.

The fissile 30-series cylinders for HALED Demonstration are limited to 5 weight percent 235U. The NCS analyses relied upon to demonstrate compliance with the double contingency principle and the 10 CFR 70.61 performance requirements are extremely conservative. Each 30-series cylinder is modeled with significantly more fissile material than what can be present in actual circumstances. In addition, the base model also eliminates spacing between cylinders in the storage array that could ACO 24-0035 Page 7 of 16 contribute to isolation of adjacent cylinders. This conservative cylinder base model is then employed to evaluate upset conditions for 30-series cylinder operations. Upset conditions for the 30-series cylinder operations are also very conservative including; 1) stacked cylinder upsets modeled a large number of stacked cylinders; 2) water intrusion upsets modeled a greater than credible amount of water entering an impacted cylinder; 3) modeled moderation/reflection conditions were extreme, including fully flooded storage array conditions; 4) interaction upsets modeled two failures instead of one; and finally 5) enrichment upsets modeled beyond credible amounts of higher enriched material in the impacted cylinder. Given the extreme configurations evaluated, and large number of failures required to achieve the configurations, the risk of criticality for 30-series cylinder

handling, storage, and transportation is well beyond extremely unlikely. Therefore, the conservative evaluation of upsets, combined with the robust controls for 30-series cylinder operations, supports a conclusion of a trivial risk of criticality.

The increased vehicular and pedestrian traffic necessary to support CAAS maintenance and calibration requirements would increase the likelihood for fire and impact events for 30-series cylinder operations such that workers would be at a higher risk for injury and exposure relative to the minimal mitigative value produced by the presence of CAAS. Therefore, CAAS coverage is not necessary for the handling, storage, and transportation of 30-series cylinders.

Further discussions related to this exemption for the HALEU Demonstration are documented within Section 3.10.4.1.1 ofLA-3605-0003A (Enclosure 5).

B. CAAS Out of Service Exemption ACO requests an exemption from the requirements of 10 CFR 70.24 for circumstances in which the CAAS is out of service and compensatory measures are in place. Without this exemption, it is necessary for ACO to provide 10 CFR 70.50(b) notifications when CAAS is taken out of service, such as to support annual CAAS surveillances or other maintenance activities that require a CAAS outage.

Pursuant to 10 CFR 70.50(b)(2), on August 1, 2023, ACO notified the NRC of a pre-planned outage of the CAAS to perform periodic testing (NRC Event#56647).

Compensatory measures were implemented in accordance with Section 5.4.4 of the License Application.

These measures included the following: evacuation of non-essential personnel from the area of concern and the immediate evacuation zone (IEZ) before removing CAAS equipment from service, limiting access into the area; restricting fissile material movement; and the use of Personal Alarming Dosimeters for personnel that must access the area during the CAAS outage. These measures were implemented until CAAS ACO 24-0035 Page 8 of 16 coverage was verified to be operational, and the CAAS declared operable in accordance with approve<l'plant procedures. The event was terminated after completion of the routine periodic testing and pursuant to 10 CFR 70.50( c )(2) on September 28, 2023, ACO submitted (ACO 23-0068) the required 60-day written event report to the NRC.

As evaluated and discussed in Section 5.3.4.4 of the previously approved NUREG-1851, Safety Evaluation Report for the American Centrifuge Plant [ML062700087] dated September 2006, in the unlikely event CAAS coverage is lost, appropriate compensatory measures will be imposed. The applicant stated in Section 5.4.4 of the License Application that it would plan and document compensatory measures as part of off-normal operation procedures before initiation of operations. Section 5.4.4 of the License Application also indicates that these may include equipment shutdown, limiting access, or halting movement of uranium-bearing material. These compensatory measures satisfy the acceptance criteria in Section 5.4.3.4.3(7) of NUREG-1520.

These measures are consistent with usual industry practice and are therefore acceptable to the NRC staff.

The exemption being proposed in Section 1.2.5 of the License Application states as follows:

In the event CAAS or its associated equipment is out of service, ACO is exempt from the requirements of 10 CFR 70.24 provided that compensatory measures are employed and remain effective until the CAAS has been restored to service. Plant procedures provide for compensatory measures which include limiting access, halting movement of fissile material, and use of Personal Alarming Dosimeters for personnel who access the area during a CAAS outage. The Personal Alarming Dosimeters used to augment the installed CAAS are evaluated against ANSI/ANS-8.3-1997 and the criteria for use shall be specified in procedures.

Further discussions related to this exemption for the ACP and for HALEU Demonstration are documented within Section 3.10.6.1.2 ofLA-3605-0003 (Enclosure 4).

Therefore, ACO proposes that the proceduralized compensatory measures described in the License Application support documentation to "justify an acceptable alternative to meet the regulations" and provide equivalent risk reduction to the 10 CFR 70.24 compliant CAAS when implemented prior to CAAS outages. Based on the above discussion, an exemption from 10 CFR 70.24 for circumstances in which the CAAS is out of service and compensatory measures are in place is justified. Without this exemption, it is necessary for ACO to provide 10 CFR 70.50(b) notifications when CAAS is taken out of service, resulting in unnecessary administrative burden. Additionally, an exemption for CAAS outages when compensatory measures are employed is consistent with Section 4.4.2 of ANSI/ANS-8.3-1997.

ACO 24-0035 Page 9 of 16 NRC has granted a similar exemption to other fuel cycle* facilities. Reporting planned CAAS outages results in an administrative burden to both ACO and the NRC with no added value. This exemption will not alter reporting requirements for inadvertent criticality events.

C. Non-Fissile Material Operations NFMOs are controlled such that the enrichment and/or inventory are maintained below the enrichment or fissile mass necessary for criticality. The determination of which operations are fissile versus which operations are non-fissile are made by NCS and includes consideration of normal and credible abnormal upset conditions to ensure the enrichment and/or inventory are maintained below 1 weight percent 235U or below 100 grams 235U. Section 5.4.4 of the License Application and Section 5.3.4.4 of the previously approved NUREG-1851, state that NFMOs and areas with less than 700 grams 235U do not require CAAS under 10 CFR 70.24(a).

NCS has determined that these operations cannot be made critical under normal and credible abnormal conditions at the ACP. For items containing less than 100 grams 235U, this mass is a factor of 7 below the minimum critical mass for uranium enriched to 100 weight percent 235U, regardless of whether the material is optimally moderated and fully reflected. HALEU Demonstration is limited to less than 20 weight percent 235U, providing additional conservatism.

Conversely, CAAS false alarms may result in process disruption and injury to personnel.

Maintaining a CAAS in NFMO areas may also result in increased occupational exposure to personnel and increased foot traffic in those areas. ANSI/ANS-8.3-1997, Section 4.1.3 states that "the purpose of an alarm system is to reduce the risk to personnel" and that "evaluation of the overall risk should recognize that hazards may result from false alarms and subsequent sudden interruption of operations and relocation of personnel." Since these NFMOs cannot be made critical, there is no tangible risk benefit to requiring a CAAS for NFMO areas. Therefore, this exemption is being proposed in Section 1.2.5 of the License Application for completeness and states as follows:

Non-fissile material operations do not require CAAS coverage. ACO has established a threshold of 1 weight percent or higher enriched 235U and 100 g or more of 235U for determining what evaluation for NCS considerations of planned operations must be performed. Operations in which the uranium enrichment is less than 1 weight percent or an inventory of less than 100 g 235U 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 is made by NCS and may be contained within a NCSE or as a separate document. The determination of an operation being non-fissile includes normal and credible abnormal upset conditions to ensure the enrichment and/or ACO 24-0035 Page 10 of 16 inventory are maintained below 1 weight percent 235D or below 100 g 235D. This 100 g 235D mass is a factor of 7 below the minimum critical mass, regardless of whether the material is optimally moderated and fully reflected for uranium enriched to 100 weight percent 235D. Based on this, the value is sufficiently low to use as a threshold limit for exemption from CAAS_ coverage.

Further discussions related to this exemption for the ACP and for HALED Demonstration are documented within Section 3.10.6.1.3 ofLA-3605-0003 (Enclosure 4).

Similar exemptions have previously been approved by the NRC and are actively used by other fuel cycle facilities for CAAS [

References:

ML19088A101, ML093441396, ML103410249].

D. Operations for Which a Criticality Accident is Not Credible As stated within Section 5.4.4 of the License Application and confirmed in Section 5.3.4.4 of the previously approved NDREG-1851, CAAS is not required for areas in which an NCSE has evaluated the fissile material operation and determined that a criticality accident is not credible. The basis for incredibility requires that the inventory of the area be less than 700 grams 235D at a minimum. The 700 grams 235D inventory is consistent with the subcritical mass limit specified in Table 1 of ANSI/ANS-8.1-2014.

Section 5.4.4 of the License Application and Section 5.3.4.4 ofNDREG-1851 further states that areas with less than 700 grams 235D do not require CAAS under 10 CFR 70.24(a).

Therefore, this exemption is being proposed in Section 1.2.5 of the License Application for completeness and states as follows:

CAAS coverage is not required for an area in which an NCSE has evaluated the fissile material operation and determined that a criticality accident is not credible. The basis for incredibility requires that the inventory of the area be less than 700 g 235D at a minimum. The 700 g 235D inventory is consistent with the subcritical mass limit specified in Table 1 of ANSI/ANS-8.1-2014. The conclusion that a criticality in the area is not credible must be documented in an NCSE per Section 5.4.4 of this License Application.

Further discussions related to this exemption for the ACP and for HALED Demonstration are documented within Section 3.10.6.1.4 ofLA-3605-0003 (Enclosure 4).

Similar exemptions have previously been approved by the NRC and are actively used by other fuel cycle facilities for areas with less than 700 grams 235U. [

References:

ML19088Al01, ML093441396, ML103410249].

I ACO24-0035 Page 11 of16 E. Special Nuclear Material Packaged as Defined in 10 CFR Part 71 The transportation requirements of 10 CFR Part 71 serve to limit special nuclear materi'al quantities to shipping configurations, thus providing limitations on geometry and interaction of fissile material. Therefore, the maximum number of containers permitted.in each area shall be unlimited for low specific activity packages as defined in 49 CFR 1 73.403, and the maximum number of other fissile packages in each area ;must be limited to a criticality safety index (CSI) of 100, with at least 20 feet (6 meters) between areas.

Therefore, the NRC has granted exemptions to 10 CFR 70.24 for areas involving spec:ial nuclear material that is contained in packages defined in transportation requirements of 10 CFR Part 71.

Section 173.403 of 49 CFR de:qnes low specific activity packages as those that do not contain fissile material or are fissile exempt under 49 CFR 173.453. This material has been determined to be exempt from all requirements for the transport of fissile material, justifying the transport of unlimited numbers of such low specific activity containers. The CSI exemption is justified based on the requir_ements of 49 CFR 176. 704( e ). Based on the low inherent risk of criticality with such materials, the potential for criticality involving

• these materials is trivial and therefore will not result in undue hazards to life or property.

As discussed above for NFMO areas, false alarms may result in process disruption and injury to personnel. Additionally, granting an exemption from the requirement for a CAAS represents a savings in cost and occupational exposure required to install, maintain, and calibrate the systems.

Therefore~ this exemption is being proposed in Section 1.2.5 of the License Application for completeness and states as follows:

CAAS coverage is not required for storage areas in which the only special nuclear material present is contained in packages as defined in 10 CFR Part 71 or specifically exempt according to 10 CFR 71.15. The maximum number of containers permitted in each area shall be unlimited for low specific activity packages, and the maximum number of fissile packages in each area must be limited to a criticality safety index (CSI) of 100, with at least 20 feet (6 meters) between areas. The transportation requirements serve to limit special nuclear material quantities to shipping configurations, thus providing limitations on geometry and interaction of fissile material. Therefore, the potential for criticality involving these materials is trivial and will not result in undue hazards to life or property. The increased vehicular and pedestrian traffic necessary to support CAAS maintenance and calibration requirements would increase the likelihood for fire and impact events for these areas such that workers would be at a higher risk for injury and exposure relative to the minimal mitigative value produced by the presence of CAAS. Therefore, CAAS coverage is not required for storage areas in which the only special nuclear material present is contained ACO24-0035 Page 12 of 16 in packages as defined in 10 CPR Part 71 or specifically exempt according to 10 CPR 71.15.

Further discussions related to this exemption for the ACP and for HALED Demonstration are documented within Section 3.10.6.1.5 ofLA-3605-0003 (Enclosure 4).

Similar exemptions have previously been approved by the NRC and are actively used by other fuel cycle facilities for material that is contained in packages defined in 10 CFR Part 71 or specifically exempt according to 10 CPR 71.15 [

References:

ML19088A101, ML093441396, and ML103410249).

Exemptions for NFMO areas; operations for which a criticality is not credible; areas where material is contained in packages defined in 10 CFR Part 71; and the handling, storage, and transportation of fissile 30-series cylinders are warranted because the risk of criticality is trivial and there is no risk benefit to requiring a CAAS consistent with Sections 4.1.1 and 4.1.3 of

, ANSI/ANS-8.3-1997.

Therefore, based upon the above discussions, the requested exemptions are allowed by 10 C:fR 70.17(a) which provides for exemptions that are authorized by law and will not endanger life'or property or the common defense and security and are otherwise in the public interest. The potential for criticality involving the operations listed above is very low (trivial) and will not endanger life or property. Additionally, the proposed exceptions do not impact the common defense and

• security. The requested exemptions are based on the *Operating experience of similarly licensed fuel cycle facilities employing similar compensatory measures, and accepted consensus standards.

I Implementation of these proposed changes will not impact the design function, or method of performance or controlling design functions, structures, systems, and components, nor will the proposed changes decrease the effectiveness of any program or plan contained in the License Applications and Supporting Documents. Moreover, the proposed changes will not change the assumptions, or change, degrade or prevent actions described or assumed in accident sequences evaluated and described in the Integrated Safety Analysis (ISA) Summary/Addendum 1 of the ISA Summary for HALED Demonstration, nor will any Items Relied on for Safety (IROFS) be affected. Therefore, no credible accident sequences could exceed the 10 CFR 70.61 performance requirements and the potential exposure to the general public is anticipated to be negligible.

Environmental Considerations ACO proposes that there are no significant environmental impacts associated with the changes proposed in this amendment request. The proposed changes do not meet the criteria in 10 CPR 5 l.60(b )(2) since they do not involve a significant expansion of the site, a significant change **tn the types of effluents, a significant increase in the amounts of effluents, a significant increase in individual or cumulative occupational radiation exposure, or a significant increase in the potential for or consequences from radiological accidents.

ACO 24-0035 Page 13 of 16 Additionally, this amendment request does not impact the transportation related to the receipt of feed material as previously evaluated for the HALEU Demonstration project as currently described within the License Application and Supporting Documents. Therefore, it is believed that the previously documented Environmental Assessment for the Proposed Amendment of US. Nuclear Regulatory Commission License Number SNM-2011 for the American Centrifuge Plant in Piketon, Ohio [ML21085A705] dated June 2021 remains bounded. The 2021 Environmental Assessment states in part:

NUREG-1834, Environmental Impact Statement for the Proposed American Centrifuge Plant in Piketon, Ohio [ML06125013 l Volume 1 and ML061250101 Volume 2] date published April 2006, estimated that 1,100 feed shipments would be shipped annually to the [ commercial] ACP, evaluated that number of shipments, and determined transportation impacts from operation to be small. NUREG-1834 also estimated there would be 2,286 truckloads of construction-related material during the first five years of the [ commercial ACP] license, evaluated that number of truckloads, and determined transportation impacts from construction to be small to moderate.

The potential transportation impacts during operation of the HALEU cascade would be due to feed shipments. The HALEU feed material would be shipped in U.S. origin 30-B series cylinders that have a 2.5-ton capacity. ACO would receive a very small fraction of the estimated feed shipments for the commercial ACP.

The NRC staff evaluated -the small number of additional feed shipments when compared to the 1,100 feed shipments evaluated for the full [commercial] ACP, as well as daily vehicular traffic, and concludes that there would not be a significant impact due to transportation a~tivities from the proposed action.

For the anticipated license amendment request, seeking operation of the HALEU cascade for an additional period of up to 10-years, ACO estimates that three additional shipments of feed material per year would be necessary to meet the expected level of production. Transport ofHALEU product is not expected to occur during the period of continued operation. Considering this small number of feed shipments, when compared to the daily vehicular traffic and the larger number of feed material shipments anticipated for the ACP, the NRC staff does not anticipate a significant impact due to transportation activities during the period of continued operation.

Consequently, a separate supplement to the Environmental Report is not being submitted.

Significance Determination for Proposed Conforming Changes Enclosuve 1 ACO 24-0035 Page 14 of 16 ACO has reviewed the proposed changes to provide exemptions from the 10 CPR 70.24 CAAS requirements~and provides the following Significance Determination:

1. No significant change to any conditions to the License.

The proposed changes to the CAAS exemptions are not prohibited by 10 CPR Part 70, license condition, or order. The Lic,ensee's initial exemption from criticality monitoring requirements of 10 CPR 70.24 for UF6 cylinder storage yards was authorized by the NRC on September 11, 2006, with the issuance ofNUREG-1851. 10 CPR 70.17 allows exemption to be granted when such relief is authorized by law, does not endanger life, nor property, nor the common defense and security, and is otherwise in the public interest. Materials License SNM-2011, specifically Condition 12, grants special authorizations and exemptions identified in Section 1.2.5 of the ACP License Application; thereby, will be updated by the NRC to reflect the newest approval of these requested exemptions. No other License Conditions are impacted by this amendment.

2. No significant increase in the probability of occurrence or consequences of previously evaluated accidents.

The proposed changes to the CAAS exemptions do not remove or alter an IROPS that is listed in the ISA Summary or Addendum 1 of the ISA Summary. Nor do the proposed changes exceed the performance requirements of 10 CPR 70.61; therefore, there is no significant increase in the probability of occurrence or consequences of the previously evaluated accidents.

3. No new or different type of accident.

The accident of concern is a criticality event. The CAAS is not credited as an IROPS; therefore, the exemptions from 10 CPR 70.24 CAAS monitoring do not create new types of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of 10 CPR 70.61 and that have not previously been described in the ISA Summary/Addendum 1 of the ISA Summary.

4. No significant reduction in the margins of safety.

The proposed changes to CAAS exemptions do not decrease the margin of safety associated with any IROPS being credited to ensure the performance requirements of 10 CPR 70.61 are met.

5. No significant decrease in the effectiveness of any programs or plans contained in the licensing documents.

No changes are required for ACO's classified matter protection security plans in Oak Ridge, 1N; Piketon, OH; or Bethesda, MD under this license amendment. Therefore, the ACO 24-0035 Page 15 of 16 proposed changes to the CAAS exemptions will not decrease the overall level of security performance needed to protect against the loss or compromise of classified matter, while in use, storage, or in transit. The control of classified storage areas or vaults, training of classifiers,,documentation of classified matter, is maintained at the same levels.

Additionally, no changes are required for security plan SEC-18-0002, American Centrifuge Operating, LLC (ACO) Information System Security Plan (ISSP) for Oak Ridge, TN; Piketon, OH; and Bethesda, MD, which provides for the protection of cyber systems, maintaining the necessary computer security requirements at the same level as previously approved by the NRC. Therefore, the security plans in Oak Ridge, TN; Piketon, OH; and Bethesda, MD remain compliant with 10 CFR Part 95 requirements.

No changes are required for the Fundamental Nuclear Material Control Plans (FNMCP) supporting the ACP (NR-3605-0005) and HALEU Demonstration (NR-3605-0005A).

Therefore, the proposed changes to the CAAS exemptions will have no effect on the FNMCPs meeting the applicable requirements of 10 CFR Parts 70 and 74. Likewise, the proposed changes do not affect the function or process to control nuclear material as described within the FNMCPs.

The proposed changes to NR-3605-0008 related to the CAAS exemptions do not change any regulatory requirements or commitments within the Emergency Management program, but instead provide clarification regarding the requirements for exemptions to CAAS coverage and where approved CAAS exemptions can be found in the License Application.

These proposed changes do not change any emergency preparedness requirements, capabilities, methods, resources, or implementation for the commercial ACP Emergency Management program. Additionally, the proposed changes will not decrease the abilities of the U.S. Department of Energy's reservation responses organization to mitigate accident cpnsequences or reasonably assure the adequate protection of the health and safety of the off-site and on-site personnel in the event of an emergency. Therefore, the proposed changes do not decrease the effectiveness of the Emergency Plan because the commercial ACP Emergency Plan, as changed, still meets NRC requirements of 10 CFR 70.22(i)(3).

Additionally, DAC-3901-0005, Evaluation of No Need for an Emergency Plan for the HALEU Demonstration, provides the evaluation stipulated in 10 CFR 70.22(i)(l)(i) to demonstrate that no Emergency Plan is needed for the HALEU Demonstration and has been written with consideration of the format and content guide provided in NUREG 1520, Section 8.4.3.2, Evaluation that No Emergency Plan is Required. The evaluation satisfies the 10 CFR 70.22(i)(l )(i) requirement to demonstrate "that the maximum dose to a member of the public offsite due to a release of radioactive materials would not exceed 1 rem effective dose equivalent or an intake of 2 milligrams of soluble uranium." No changes are warranted for this evaluation based upon the proposed exemptions being requested.

The proposed changes to the CAAS exemptions do not result in a change to the Quality Assurance Program Description (NR-3605-0003); thereby, do not represent a relaxation of a requirement of Quality Assurance Program Description.

ACO 24-0035 Page 16 of 16 Based on the above, the proposed changes will *not result in a decrease in the effectiveness of the Security Programs/Plans, FNMCPs, Emergency Plan, or the Quality Assurance Program Description contained in the licensing documents.

6. The proposed change does not result in undue risk to: 1) public health and safety; 2) common defense and security; and 3) the environment.

The proposed changes to the CAAS exemptions do not change the response to accidents or events associated with licensed material. There will be no generation or increase in hazardous material quantities such that it impacts public health and safety. The proposed changes h~ve no impact to the plant boundary protection, documentation of patrols, performance of rounds, or training of protective force personnel. The proposed changes will not increase the likelihood classified matter or special nuclear material will be accessible to unauthorized personnel.

Physical protection methods for special nuclear material remain unchanged. Therefore, the proposed changes do not result in undue risk to public health and safety, the environment, or to the common defense and security.

7. There is no change in the type or significant increases in the amounts of any effluents that may be released off-site.

The proposed changes to the CAAS exemptions do not result in any new or unusual sources of hazardous substances, hazardous waste, or new waste streams that could be generated or used in unacceptable levels that exceed applicable regulatory requirements. In addition, there is no change in the type or significant increases in the amounts of any effluents that may be released off-site. The amount of material is much less than currently evaluated.

8. There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes to the CAAS exemptions will not increase radiological or chemical releases beyond applicable regulatory limits (10 CFR 70.61) and will not create any new or unusual sources of radioactive waste. Likewise, the proposed changes will not result in significant increase in individual or cumulative occupational radiation exposure. CAAS exemptions are requested for operations with trivial or no chance of criticality, such that dose or exposure from a criticality accident is not increased.

9. There is no significant construction impact.

The proposed changes are administrative in nature and have no known impacts to any planned construction activities at Piketon, Ohio; Oak Ridge, Tennessee; or Bethesda, Maryland. Thus, there are no foreseen environmental concerns or impacts.

of ACO 24-0035 Proposed Changes to LA-3605-0001, License Application for the American Centrifuge Plant Information Contained Within Does Not Contain Export Controlled Information Reviewing Official:

Lori Hawk/ ACO Date:

04/26/2024

LA-3605-0001 License Application for the American Centrifuge Plant in Piketon, Ohio Docket No. 70-7004 Proposed Change Information Contained Within Does Not Contain Export Controlled Information Reviewing Official:

Lori Hawk/ ACO

• Date:

04/26/2024 May2024

FOR INFORMATION ONLY License Application for the American Centrifuge Plant Proposed Change May 2024 The design of the plant complies with the performance requirements of 10 CFR 70.61, the Baseline Design Criteria specified in 10 CFR 70.64(a) and the defense-in-depth requirements contained in 10 CFR 70.64(b).

1.1.3 Primary Facilities Description Primary facilities are those buildings/facilities or areas that could potentially contain licensed material in quantities that result in consequences that exceed the performance criteria defined in 10 CFR 70.61 resulting from credible accidents or directly controls a primary facility.

The primary facilities directly involved in the enrichment process are the X-2232C Interconnecting Process Piping (IPP), X-3001 Process Building; X-3002 Process Building; X-3012 Process Support Building; X-3344 Customer Services Building; X-3346 Feed and Withdrawal Building; and X-3346A Feed and Product Shipping and Receiving Building. Other buildings and areas that provide direct support functions to the enrichment process are the X-7725 Recycle/Assembly Building; X-7726 Centrifuge Training and Test Facility; X-7727H Interplant Transfer Corridor; X-745G-2 Cylinder Storage Yard; X-745H (future) Cylinder Storage Yard, and X-7746S, X-7746W Cylinder Storage Yards and Intraplant Roadways. These buildings and areas are where, special nuclear material and hazardous material can be found and are considered to be the primary facilities in their functional support of the uranium enrichment process. A description of the primary facilities and their function is provided in the following sub-sections and are listed and briefly described in Table 1.1-1. An overall depiction of the enrichment processes is provided in Figure 1.1.3-f located in Appendix E.

ACO's long-term goal is to resume commercial enrichment production consistent with market demand. The ACP design is modular, with the basic building block of enrichment capacity being a cascade of centrifuges. Modular deployment would accommodate market demand on a scalable, economical gradation. The Fire Safety Program will be implemented to support the modular deployment, such that the fire protection systems/services are in place when needed.

The next phase of enrichment production includes the deployment of a cascade of 16 centrifuges to demonstrate production of high-assay, low-enriched uranium (HALEU) fuel for advanced reactors. The primary building/facilities directly involved in HALEU Demonstration are the X-3001 Process Building, X-3012 Process Support Building, X-7725 Recycle/Assembly Building, X-7726 Centrifuge Training and Test Facility, and X-7727H Interplant Transfer Corridor. It is also noted that HALEU Demonstration does not involve or include the use of any liquid UF6 handling operation or those facilities.

1.1.3.1 X-3001 and X-3002 Process Buildings The initial deployment of the ACP includes two process buildings, which are located in the southwest quadrant of the DOE reservation: X-3001 and X-3002. The primary purpose of the process buildings is to house the centrifuges and support systems necessary to perform the actual enrichment process. Both buildings are similar in construction, layout, and design. Each building is approximately 416 feet (ft) by 730 ft (approximately 304,000 square feet [ft2]) and has a large high bay process area and two utility areas. The height of each building is approximately 87 ft in the high bay area and 49 ft in the utility areas. The nearest reservation boundary is 2,606 ft to the west of the X-3001 building. Figure 1.1-3 (located in Appendix B) depicts the typical equipment 1-4

License Application/or the American Centrifuge Plant Proposed Change May 2024 and process flow for the X-3001 and X-3002 buildings. Figures l.l.3. 1-1, 1.1.3.1 -2, 1.1.3.1-3, and 1.1.3.1-4 (located in Appendix E) also depict the equipment layout for the X-300 l and X-3002 buildings.

At the north and south ends of X-3001 and X-3002 buildings are equipment/utility bays and mezzanines where auxiliary equipment is housed. Items in these areas consist of heating and ventilation equipment, cooling water pumps, vacuum pumps, electrical switchgear, and standby electrical equipment (i.e., diesel generators, battery rooms, and uninterruptible power supply [UPS]

systems). Building vents for the purge and evacuation vacuum systems are also located in the buildings.

The vents are monitored and are permitted through the Ohio Environmental Protection Agency (OEPA).

The east side of the X-3001 building is connected to the X-3012 building, which is connected to the west side of the X-3002 building. The X-7727H corridor is connected to the west side of the X-3001 building. The X-2232C piping connects to the southwest comer of the X-3001 building at a valve house where it both enters and exits the building. The connection of the X-2232C piping exits the east side of the X-3001 building and enters and exits the X-3002 building on the west side through a valve house as well.

The centrifuges are installed in the high bay area in a cascade arrangement. The cascades are supplied UF6 feed from a header from the Feed Area in the X-3346 building. The centrifuges in each cascade are grouped into stages that are connected in series. The feed, product, and tails lines to and from each centrifuge within a stage connect into stage headers that convey the UF 6 streams between stages. The depleted material from the bottom stage is piped through the X-2232C piping to the X-3346 building Withdrawal Area to be withdrawn as tails. The enriched material from the top stage is piped through the X-2232C piping to the X-3346 building Withdrawal Area to be withdrawn as product. For commercial ACP operations the cascade enrichment is normally less than 5.5 wt. percent 235U, but enrichment levels up to 10 wt. percent 235U are allowable.

The HALEU Demonstration cascade utilizes a similar centrifuge design to that used for the Lead Cascade. The equipment necessary to perform the enrichment process is in the X-3001 Process Building and consists of product and tails withdrawal system, UF 6 cylinders, centrifuges, and supporting units. The product and tails withdrawal systems use three cold boxes. NaF traps are used for additional withdrawal capacity during dumping. A 30B Uf 6 cylinder is used for the feed material. Centrifuges and supporting units are placed in the Train 3 area of the X-3001. For further plant and process specifics related to the HAL EU Demonstration Program, refer to LA-3605-0003A, Addendum 1 of the Integrated Safety Analysis Summary (or the American Centrifuge Plant-HALEU Demonstration (Reference 7).

1.1.3.2 X-3012 Process Support Building The X-3012 houses the operational area, maintenance area, and the transfer aisleway that services the X-3002 building. The X-3012 building is located between the X-300 l and X-3002 buildings. The X-3012 building, which is approximately 201 ft by 240 ft at grade level, has a ground floor area of approximately 48,000 ft2, and has a total covered floor space area of approximately 56,200 ft2, which includes the ground floor and two mezzanine areas. The transfer aisle way between the X-300 l and X-3002 and through the X-3012 building measures 30 ft wide by approximately 59 ft high by 200 ft long and divides the building into north and south sections.

The north section is approximately 17 ft high and contains the operational area. The south section 1-5

license Application/or the American Centrifuge Plant Proposed Change May 2024 Other areas of the ACP are provided with HV AC or only heating and ventilation, depending on the location and function of the area or facility. Supplemental heat can be provided in any ACP facility using portable electric heaters should the RHW be out of service or outside weather conditions dictate the need.

1.1.5.6.5 Criticality Accident Alarm System The primary radiation alarm system is the CAAS designed to detect a nuclear criticality and provide audible and visual alarms that will alert personnel to evacuate the immediate area.

The CAAS is described in Section 5.4.4 of this license application. The CAAS coverage areas are identified on plant drawings and controls are established to preclude special nuclear material from areas where coverage is not provided.

Per IO CFR 70.24, CAAS is required in each area where special nuclear material is handled, used, or stored by a licensee authorized to possess more than 700 g of 235U.

AGP-ACO primary facilities that handle 235U in quantities exceeding 700_g and enrichment levels greater than or equal to 1 weight percent have CAAS coverage unless the NRC has granted an exemption from the 10 CFR 70.24 CAAS requirements. e~teept the UF6 eylinder storage )'ards. Att-eg xemptions from -fet:

the ~

eylinder storage )'ards has been reEJHested requirements of IO CFR 70.24 are documented in Section 1.2.5 of this -b!icense A~pplication. C)1linders are mo'ved betweeA the YarioHs bHildiAgs with the material iA a solid state on approYed aAd defiAed roHtes HsiAg speeifieally desigAed eEJHipment iA aeeordaAee 1Nith approved proeedHres that are eo*,*ered b)' CAAS.

OperatioAs iAvoh*ing fissile material are e't*alHated for NHelear Critieality Safety (JI-JCS) eonsiderations prior to initiation. The need for CAAS eoverage is eonsiderea aHring the e111alHation proeess. Co11erage is proYiaea, Hnless it is determiAed that eo*,*erage is not reEJHired ana the finding is doeHrnented in a },JCS 61,*alHation. Per IQ CFR 70.24, CAAS is reEJHired in eaeh area \\11*here threshold EJHantities (e.g., more than ?QQ grams of ~ U) of special AHelear material are handled, Hsed, or stored. The CAAS eo111erage areas are identified on plant drawings, and eontrols are established to preelHde speeial nHelear material from areas *n*here eoYerage is not pro,.*ided.

1.1.5.6.6 Portable Gulpers A portable gulper system is used for localized exhaust on applications like small-scale maintenance tasks. The gulper inlet duct or hose is placed near the work area. Any escaping airborne contamination is removed from the source and passes through the duct or hose and into the filter bank, where, depending on the operation, gases are neutralized and the particulates are removed. The resultant exhaust is clean air that is typically discharged into the work area.

1.1.5.7 Centrifuge Assembly and Movement Systems 1.1.5.7.1 Centrifuge Assembly The centrifuges are assembled in the X-7725 building and/or the X-7726 faci lity assembly stands. Parts for the centrifuge assembly are received at these locations. Secure facilities are available to receive and store the classified parts, as well as other components of the centrifuges.

1-27

license Application for the American Centrifuge Plant Proposed Change May 2024 and State EPA as required by the Superfund Amendments Reauthorization Act (SARA Sections 312 and 313).

The Licensee complies with requirements for generators of hazardous and mixed waste.

The State of Ohio has adopted a federal conditional exemption from the hazardous waste rules that is available under 40 CFR Part 266, Subpart N (OAC 3745-266).

1.1.7 Roadways Two major four-lane highways service the DOE reservation: U.S. Route 23, traversing north-south, and State Route 32/124, traversing east-west.

The reservation is situated approximately three and one half miles from the intersection of U.S. Route 23 and State Route 32/124. There are five major access roads, which connect Perimeter Road to adjoining roads outside the DOE reservation. The major one is the West Access Road (Principal Access Road) from U.S. Route 23, which lies approximately one mile west of Perimeter Road. The North Access Road, which connects to U.S. Route 32 is approximately three miles to the north. The East and South Access Roads connect to secondary county roads. There is also a construction entrance road on the southwest corner of the reservation, which ties into Perimeter Road. This road was used during the original site and facility construction periods. Vehicle traffic access to the Perimeter Road is open to the public but can be shut down as necessary for safety and security concerns, or in support of reservation activities. Service roads throughout the reservation connect to the Perimeter Road with access to the ACP controlled through security portals. The reservation roadways are depicted in Figures 1.1-1 (located in Appendix B) and 1.1-2.

1.1.8 Phased Modular Expansion Plan for the American Centrifuge Plant It is the intent of ACO to deploy portions of the ACP in a modular fashion to accommodate market demand on a scalable, economical gradation. This modular deployment may encompass utilization of cascades of Low Enriched Uranium (LEU) production for LEU customer product or feed material into HALEU cascades. The ratio of LEU cascades to HALEU cascades would be approximately 6 to l.

1.1.8.1 High Assay Low Enriched Uranium Demonstration The HALEU Demonstration cascade utilizes a similar centrifuge design to that used for the Lead Cascade. The equipment necessary to perform the enrichment process is in the X-3001 Process Building and consists of product and tails withdrawal system, Uf 6 cylinders, centrifuges, and supporting systems. The product and tails withdrawal systems use three cold boxes. NaF traps are used for additional withdra~al capacity during dumping. A 30B UF6 cylinder is used for the feed material. Centrifuges and supporting units are placed in the Train 3 area of the X-300 I building. For further plant and process specifics related to the HALEU Demonstration Program, refer to LA-3605-0003A, Addendum 1 of the Integrated Safety Analysis Summary for the American Centrifuge Plant - HALEU Demonstration (Reference 7).

In support of this HALEU Demonstration Program and NRC Materials License (SNM-2011) Condition 23, DOE amended (Amendment l) the Appendix 1 Lease Agreement between the U.S. Department of Energy and United States Enrichment Corporation for the Gas Centrifuge 1-30

FOR INFORMATION ONLY License Application for the American Centrifuge Plant Proposed Change May 2024 authorized an increase in the possession limits beyond those approved on June 11, 2021 (SNM-2011, Amendment 19).

3) Within the ACP Operations, the Licensee will provide a minimum 60-day notice to the NRC prior to initial customer product withdrawal of licensed material exceeding 5 wt.

percent 235U enrichment. This notice will identify the necessary equipment and operational changes to support customer product withdrawal, storage, processing, and shipment for these assays.

1.2:s Special Exemptions or Special Authorizations The following exemption to the applicable 10 CFR Part 20 requirements are identified in Section 4.8 of this license application:

UF6 feed, product, and depleted uranium cylinders, which are routinely transported inside the DOE reservation boundary between ACP locations and/or storage areas at the ACP, are readily identifiable due to their size and unique construction and are not routinely labeled as radioactive material. Qualified radiological workers attend UF6 cylinders during movement.

Containers located in Restricted Areas r within the ACP are exempt from container labeling requirements of 10 CFR 20.1904, as it is deemed impractical to label each and every container. In such areas, one sign stating that every container may contain radioactive material will be posted. By procedure, when containers are to be removed from contaminated or potentially contaminated areas, a survey is performed to ensure that contamination is not spread around the reservation.

In lieu of the requirements of 10 CFR 20.1601(a), each High Radiation Area with a radiation reading greater than 0.1 Roentgen Equivalent Man per hour (REM/hour) at 30-centimeters (cm) but less than 1 REM/hour at 30 cm is posted Caution, High Radiation Area and entrance into the area shall be controlled by an RWP. Physical and administrative controls to prevent inadvertent or unauthorized access to High and Very High Radiation Areas are maintained. The on-site radiological impacts from the proposed exemptions to the requirements of 10 CFR 20.1904 and 20.1601 would be minimal and are consistent with previously approved exemptions found in the GDP certification. Moreover, pursuant to the regulations in 10 CFR 20.2301, the requested exemption is authorized by law and would not result in undue hazard to life or property.

The following exemption from the applicable 10 CFR 70.50 reporting requirement is identified in Section 11.6.3 of this license application:

The 10 CFR 70.50( c )(2) reporting criteria require that the ACP submit a written follow-up report within 30 days of the initial report required by 10 CFR 70.50 (a) or (b) or by 10 CFR 70.74 and Appendix A of Part 70. In lieu of the 30-day requirement described in 10 CFR 70.50(c)(2), NRC approval to submit the required written reports within 60 days of the initial notifications is hereby requested.

1-63

FOR INFORMATION ONLY License Application for the American Centrifuge Plant Proposed Change May2024 10 CPR 70.17 allows the Commission, upon application of any interested person or upon its own initiative, to grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The requested exemption is authorized by law because there is no statutory prohibition on extending the reporting period to 60 days.

Furthermore, granting this exemption request will not endanger life or property or the common defense and security, in that the exemption request does not relieve the ACP from other requirements contained in 10 CPR 70.50 (a) or (b) or by 10 CPR 70.74 and Appendix A of Part 70, such as I-hour, 4-hour, and 24-hour reporting requirements for defined events.

The proposed exemption would result only in written reports being submitted within the time limit currently allowed under 10 CPR 50.73 for commercial nuclear power plants.

It would be consistent with the exemption granted to the gaseous diffusion plants for reporting of events pursuant to 10 CPR 76.120(d)(2) (67 Federal Register 68699, November 12, 2002) and the exemption granted to the Lead Cascade during licensing.

This proposal allows for completion of required root cause analyses after event discovery and fewer supplemental reports, thereby reducing regulatory burden and confusion. Thus, it is clearly consistent with the public interest.

The Licensee notes that the requirements of 10 CPR 20.2201 and 20.2203 require written reports of certain events within 30 days after their occurrence. The Licensee is not requesting an exemption from these reporting requirements.

The following exemption from the requirements of 10 CPR 70.25(e) and 10 CFR40.36(d) addressing the decommissioning funding requirements is identified in Section 10.1 of this license application:

10 CPR 70.25(e) and 10 CPR 40.36(d) require, in part, that "The decommissioning funding plan must also contain a certification by the licensee that financial assurance for decommissioning has been provided in the amount of the cost estimate for decommissioning... ".

In support of HALEU Demonstration Program, as noted in Section 10.1 of this license application, DOE amended the Appendix 1 Lease Agreement between the US.

Department of Energy and United States Enrichment Corporation for the Gas Centrifuge Enrichment Plant (GCEP Lease Agreement). In the amended GCEP Lease Agreement, DOE assumes all liability for the decontamination and decommissioning of such facilities and equipment installed, and any work performed, under the HALEU Demonstration Contract with the Department including any materials or environmental hazards on the site. Therefore, exempting ACO from any financial assurance for any liability or lease turnover conditions shall be required from the Corporation (Licensee).

1-64

FOR INFORMATION ONLY License Application for the American Centrifuge Plant Proposed Change May2024 Additionally, as stated within the amended GCEP Lease Agreement, the parties agree that should any liabilities of the Corporation (Licensee) arise from or incident to the performance of work under the Demonstration Contract with the DOE shall be governed solely by such contract and any financial protection afforded to the Corporation (Licensee) as a person indemnified under the Act. DOE is expected to continue to amend the GCEP Lease in support of the HAL EU Operations contract.

The following exemption from the requirements of 10 CFR 70.25(e) and 10 CFR 40.36(d) addressing the decommissioning funding requirements is identified in Section 10.2.10.4 and the DFP of this license application:

10 CFR 70.25(e) and 10 CFR 40.36(d) require, in part, that "The decommissioning funding plan must also contain a certification by the licensee that financial assurance for decommissioning has been provided in the amount of the cost estimate for decommissioning... ".

In support of future expansion of the ACP, as noted in Section 10.2.10.4 of this license application, the financial assurance for a portion of the decommissioning costs, to include the disposition of centrifuges and UF6 tails, which constitutes a major portion of the decommissioning liability, will be provided incrementally as centrifuges are built/installed and UF6 tails generated.

Full funding for decommissioning of the facilities will be provided in the initial executed financial assurance instrument.

This exemption is justified for the following reasons: 1) It is authorized by law because there is no statutory prohibition on incremental funding of decommissioning costs. 2)

The requested exemption will not endanger life or property or the common defense and security for the following reasons: the unique modular aspects of the American Centrifuge technology allow enrichment operations to begin well before the full capacity of the plant is reached. Thus, the decommissioning liability for centrifuges and UF6 tails is incurred incrementally as more centrifuges are added to the process, until full capacity of the facility is reached; at which point the UF6 tails are generated at a relatively constant rate throughout the life of the plant. As such, requiring full funding for decommissioning liability, to include centrifuges and UF6 tails disposition, incurred over the lifetime of the plant, at the time of initial license issuance, produces an unnecessary financial burden on the licensee.

Furthermore, incremental funding of decommissioning costs, to include centrifuges and UF6 tails disposition, is justified based upon the Licensee's commitments to update the cost estimates and provide a revised funding instrument for decommissioning annually, to cover the upcoming period of operation, prior to operation at full capacity, and after full capacity has been reached to annually adjust the cost estimate for UF6 tails disposition and to adjust all other decommissioning costs periodically, and no less frequently than every three years. In addition, the relative stability of the factors, which are utilized to generate the UF6 tails volumes, allows actual inventory values to be provided for prior periods of operation and reliable estimates for the upcoming periods of operation.

The NRC has previously accepted _an incremental approach to decommissioning funding costs for the United States Enrichment Corporation's 1-65

License Application for the American Centrifuge Plant Proposed Change May 2024 operation of the GDPs. 3) Finally, granting this exemption is in the public interest for the same reasons as stated above and will facilitate deployment of gas centrifuge enrichment technology by eliminating an unnecessary financial burden on the licensee.

The following exemption~ from the requirements of IO CFR 70.24 addressing criticality monitoring ~ are identified in Section 3.1 0.6 of the ISA Summary, Section 3.10.4 of Addendum 1 of the ISA Summary for the ACP - HALEU Demonstration, and diset1ssed H1/2prepared per Section 5.4.4 of this -blicense Afil)plication. E:xemptioR is reqt:1ired for eriticality ffiORitoriRg of the Uf& C)'liRder storage yards.

10 CFR 70.24, Criticality Accident Requirements, requires that licensees authorized to possess special nuclear material in a quantity exceeding 700 g of contained 235U shall maintain in each area in which such licensed special nuclear material is handled, used, or stored, a monitoring system capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of two meters from the reacting material within one minute.

10 CFR 70.17 allows the Commission, upon application of any interested person or upon its own initiative, to grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The requested exemptions are--ffl authorized by law because there is no statutory provision prohibiting the grant of the exemption. The requested exemption~ will not endanger life or property or the common defense and security and tS-are otherwise in the public interest for the reasons discussed below.

Transportation, handling., and storage of solid UF 6 filled cylinders for the commercial ACP operations are doubly contingent. Double contingency is established by multiple controls that limit the likelihood for a solid product cylinder to be breached during transportation, handling or storage, and the likelihood for a breach to not be identified and repaired before sufficient moderation results in a criticality. Moderation control of UF6 filled cylinders is maintained by ensuring cylinder integrity through periodic cylinder inspections. If a UF 6 filled cylinder is found to be breached, the cylinder is covered within 24-hours after discovery to reduce the potential accumulation of moderating material, i.e., rainwater. This time limit ensures a corresponding heavy rainfall will not result in accumulation of sufficient amounts of water to cause a criticality. Damaged cylinders are repaired as necessary and emptied. UF6 cylinders are uniquely identified and their design requirements are controlled to further ensure cylinder integrity and reliability (i.e., Uf 6 cylinders are QL-1 components and are controlled in accordance with the Quality Assurance Program Description), and the Licensee implements onsite cylinder handling practices (i.e., requiring the use of approved equipment in accordance with approved procedures), which reduces the likelihood that a solid UF6 cylinder would be breached. These requirements are established as items relied on for safety to ensure the health and safety of the public and workers.

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license Application for the American Centrifuge Plant Proposed Change May 2024 operation of the GDPs. 3) Finally, granting this exemption is in the public interest for the same reasons as stated above and will facilitate deployment of gas centrifuge enrichment technology by eliminating an unnecessary financial burden on the licensee.

The following exemption~ from the requirements of IO CFR 70.24 addressing criticality monitoring +s-are identified in Section 3.10.6 of the ISA Summary, Section 3.10.4 of Addendum I of the ISA Summary for the ACP - HALEU Demonstration, and diseussed Hlprepared per Section 5.4.4 of this b license Agpplication. eX:emption is required for eritieality monitoring of the Uf6 eylinder storage yards.

10 CFR 70.24, Criticality Accident Requirements, requires that licensees authorized to possess special nuclear material in a quantity exceeding 700 g of contained 235U shall maintain in each area in which such licensed special nuclear material is handled, used, or stored, a monitoring system capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of two meters from the reacting material within one minute.

10 CFR 70.17 allows the Commission, upon application of any interested person or upon its own initiative, to grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The requested exemptions are--i5 authorized by law because there is no statutory provision prohibiting the grant of the exemption. The requested exemption~ will not endanger life or property or the common defense and security and

-i-&-are otherwise in the public interest for the reasons discussed below.

Transportation, handling,_ and storage of solid UF6 filled cylinders for the commercial ACP operations are doubly contingent. Double contingency is established by multiple controls that limit the likelihood for a solid product cylinder to be breached during transportation, handling or storage, and the likelihood for a breach to not be identified and repaired before sufficient moderation results in a criticality. Moderation control of UF6 filled cylinders is maintained by ensuring cylinder integrity through periodic cylinder inspections. If a UF6 filled cylinder is found to be breached, the cylinder is covered within 24-hours after discovery to reduce the potential accumulation of moderating material, i.e., rainwater. This time limit ensures a corresponding heavy rainfall will not result in accumulation of sufficient amounts of water to cause a criticality. Damaged cylinders are repaired as necessary and emptied. UF6 cylinders are uniquely identified and their design requirements are controlled to further ensure cylinder integrity and reliability (i.e., UF6 cylinders are QL-1 components and are controlled in accordance with the Quality Assurance Program Description), and the Licensee implements onsite cylinder handling practices (i.e., requiring the use of approved equipment in accordance with approved procedures), which reduces the likelihood that a solid UF6 cylinder would be breached. These requirements are established as items relied on for safety to ensure the health and safety of the public and workers.

1-66

License Application for the American Centrifuge Plant Proposed Change May 2024 The UF6 cylinders stored in storage yards are not covered by a criticality monitoring system unless those cylinders contain licensed material greater than 5.0 weight percent 235U. NCS evaluation of product cylinders of any size, configured in infinite planar arrays, containing material enriched up to 5.25 weight percent m u, has concluded that subcritical conditions are maintained. The ACP ISA has concluded that cylinders containing licensed material less than or equal to 5.0 weight percent m u cannot be involved in a criticality accident sequence that has a probability of occurrence that exceeds 5 x 10*6/year.

The frequencies of criticality events in the cylinder yards have been decreased to the Highly Unlikely range (< 10*5/year) through the establishment of preventive controls established by the ISA in accordance IO CFR 70.62. Considering the conservatism of the JSA methodology in developing the unmitigated frequency and actual historical data related to cylinder operations, the frequency values could be reduced further. This additional reduction considers the fact that during 50 years of GDP operations, only one cylinder breach has occurred due to mishandling or equipment failure. Since that occurrence, cylinder handling equipment has been redesigned and cylinder handling methods have been revised to minimize the potential for breaches to occur. Another fact not considered in the ISA is that holes with a dimension of less than one inch will self-seal such that moderating material cannot infiltrate the breach. A third factor not considered in the ISA is that enriched cylinder operations require constant use and monitoring of cylinders such that corrosion breaches in enriched cylinders are highly unlikely. Allowing for this additional reduction in frequency, the probability for a criticality event becomes incredible, therefore CAAS coverage is not necessary.

The increased vehicular and pedestrian traffic in support of CAAS maintenance and calibration requirements would cause a subsequent increased likelihood for impact events involving cylinders and there would be an increased safety risk for workers from radiation exposure due to the ongoing CAAS maintenance and calibration requirements. To meet the CAAS coverage requirements in ANSI/ ANS--8.3-1997 and the operating requirements for the ACP, enriched cylinder storage yards would require a minimum of 60 clusters. Clusters would need to be at a height of approximately 40 feet, which would require maintenance equipment and pedestrian traffic to perform testing and preventative maintenance tasks to ensure their reliability and operability.

This equipment and traffic would increase the likelihood for fire and impact events in the cylinder storage yards such that workers would be at a higher risk for injury and exposure relative to the minimal mitigative value produced by the presence of CAAS.

CAAS coverage is not required for the handling. storage. and transportation of fissile 30-series cylinders as part of HALEU Demonstration because there is a trivial risk of criticality for these operations.

The handling, storage, and transportation of fissile solid UF6 filled 30-series cylinders are based on robust passive design features that are subject to management measures and administrative controls that further ensure cylinder integrity and reliability. The passive design features for 30-series cylinder operations include the robust design of 1-67

License Application for the American Centrifuge Plant Proposed Change May 2024 the cylinders and storage array fixtures. The 30-series cylinder is fabricated of fire-resistant materials and is designed in accordance with ANSI N 14.1 (Reference 24). As such, 30-series cylinders are designed to not breach during a design basis fire and are resistant to damage from dropping, from contact with other cylinders and structures, and from corrosion. The controls, in combination with the robust design of the 30-series cylinders and the self-sealing nature of UF6 and water at the location of a credible crack or puncture, minimizes the accumulation of moderator within a fissile 30-series cylinder. The storage fixtures used in 30-series cylinder storage arrays are also made of fire-resistant materials and are designed to survive a design basis fire as well as minimize lateral movement of the stored cylinders and elevate the cylinders above the ground/floor.

The administrative controls associated with the handling, storage, and transportation of fissile 30-series cylinders support the passive design features discussed above by requiring periodic inspections of cylinders to identify damage, breaches. corrosion, and leak-related accumulation of fissile materials; limiting UF6 enrichment in the cylinders; prohibiting stacking of cylinders; establishing spacing from other fissile materials or processes; protecting the cylinders from conditions that could cause damage: and protecting the cylinders from conditions that could lead to water intrusion.

The fissile 30-series cylinders for HALEU Demonstration are limited to 5 weight percent 235U. The NCS analyses relied upon to demonstrate compliance with the double contingency principle and the 10 CFR 70.61 performance requirements are extremely conservative. Each 30-series cylinder is modeled with significantly more fissile material than what can be present in actual circumstances. In addition, the base model also eliminates spacing between cylinders in the storage array that could contribute to isolation of adjacent cylinders. This conservative cylinder base model is then employed to evaluate upset conditions for 30-series cylinder operations. Upset conditions for the 30-series cylinder operations are also very conservative including;

1) stacked cylinder upsets modeled a large number of stacked cylinders; 2) water intrusion upsets modeled a greater than credible amount of water entering an impacted cylinder; 3) modeled moderation/reflection conditions were extreme. including fully flooded storage array conditions; 4) interaction upsets modeled two failures instead of one; and finally 5) enrichment upsets modeled beyond credible amounts of higher enriched material in the impacted cylinder.

Given the extreme configurations evaluated, and large number of failures required to achieve the configurations, the risk of criticality for 30-series cylinder handling, storage, and transportation is well beyond extremely unlikely. Therefore. the conservative evaluation of upsets, combined with the robust controls for 30-series cylinder operations, supports a conclusion of a trivial risk of criticality.

The increased vehicular and pedestrian traffic necessary to support CAAS maintenance and calibration requirements would increase the likelihood for fire and impact events for 30-series cylinder operations such that workers would be at a higher risk for injury and exposure relative to the minimal mitigative value produced by the presence of 1-68

license Application for the American Centrifuge Plant Proposed Change May 2024 CAAS. Therefore. CAAS coverage is not necessary for the handling, storage. and transportation of 30-series cylinders.

In the event CAAS or its associated equipment is out of service, the licensee is exempt from the requirements of IO CFR 70.24 provided that compensatory measures are employed and remain effective until the CAAS has been restored to service. Plant procedures provide for compensatory measures which include limiting access, halting movement of fissile material, and use of Personal Alarming Dosimeters for personnel who access the area during a CAAS outage. The Personal Alarming Dosimeters used to augment the installed CAAS are evaluated against ANSI/ANS-8.3-1997 and the criteria for use shall be specified in procedures.

Non-fissile material operations do not require CAAS coverage. ACO has established a threshold of I weight percent or higher enriched m u and I 00 g or more of m u for determining what evaluation for NCS considerations of planned operations must be performed. Operations in which the uranium enrichment is less than I weight percent or an inventory ofless than I 00 g 235U 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 is made by NCS and may be contained within a NCSE or as a separate document. The determination of an operation being non-fissile includes normal and credible abnormal upset conditions to ensure the enrichment and/or inventory are maintained below 1 weight percent m u or below 100 g 235U. This 100 g m u mass is a factor of 7 below the minimum critical mass. regardless of whether the material is optimally moderated and fully reflected for uranium enriched to 100 weight percent 235U. Based on this, the value is sufficiently low to use as a threshold limit for exemption from CAAS coverage.

CAAS coverage is not required for an area in which an NCSE has evaluated the fissile material operation and determined that a criticality accident is not credible. The basis for incredibility requires that the inventory of the area be less than 700 g m u at a minimum. The 700 g 235U inventory is consistent with the subcritical mass limit specified in Table I of ANSI/ANS-8.1-2014. The conclusion that a criticality in the area is not credible must be documented in an NCSE per Section 5.4.4 of this license application.

CAAS coverage is not required for storage areas in which the only special nuclear material present is contained in packages as defined in IO CFR Part 71 or specifically exempt according to IO CFR 71.15. The maximum number of containers permitted in each area shall be unlimited for low specific activity packages, and the maximum number of fissile packages in each area must be limited to a criticality safety index (CSI) of 100. with at least 20 feet (6 meters) between areas. The transportation requirements serve to limit special nuclear material quantities to shipping configurations, thus providing limitations on geometry and interaction of fissile material. Therefore, the potential for criticality involving these materials is trivial and will not result in undue hazards to life or property. The increased vehicular and pedestrian traffic necessary to support CAAS maintenance and calibration 1-69

license Application f or the American Centrifuge Plant Proposed Change May 2024 requirements would increase the likelihood for fire and impact events for these areas such that workers would be at a higher risk for injury and exposure relative to the minimal mitigative value produced by the presence of CAAS. Therefore, CAAS coverage is not required for storage areas in which the only special nuclear material present is contained in packages as defined in IO CFR Part 71 or specifically exempt according to IO CFR 71.15.

The following exemption from the requirements of IO CFR 140.13b crediting DOE indemnity in lieu of nuclear liability insurance as discussed in Section 1.2.2 of this license application.

10 CFR 140.13b requires, that "Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear material is required to have and maintain liability insurance. The liability insurance must be the type and in the amounts the Commission considers appropriate to cover liability claims arising out of any occurrence within the United States that causes, within or outside the United States, bodily injury, sickness, disease, death, loss of or damage to property, or loss of use of property arising out of or resulting from the radioactive, toxic, explosive, or other hazardous properties of chemical compounds containing source material or special nuclear material. Proof of liability insurance must be filed with the Commission as required by § 140.15 before issuance of a license for a uranium enrichment facility under parts 40 and 70 of this chapter."

In support of this HALEU Demonstration Program, DOE amended the GCEP Lease Agreement, in which the parties agree that all work performed under the HALEU Demonstration Contract on leased premises shall be considered a permitted use; any alterations or changes to the premises pursuant to the Demonstration Contract with the DOE shall be a permitted change to the premises; and that any liabilities of the Corporation (Licensee) arising from or incident to the performance of work under the HALEU Demonstration Contract with the DOE shall be governed solely by such contract. Therefore, the Demonstration Contract exempts ACO from any financial assurance for any liability insurance during the HALEU Contract period.

In support offuture expansion of the ACP, in accordance with Section 3107 of the USEC Privatization Act, the Lease with DOE for the DOE owned facilities that will be used for the ACP includes an indemnity agreement from DOE under Section 170d of the Atomic Energy Act (AEA) for liability claims.

The Commission may, pursuant to IO CFR 140.8, upon application of any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and are otherwise in the public interest. This exemption is authorized by law because there is no statutory prohibition on crediting the DOE indemnity agreement in lieu of nuclear liability insurance. The DOE indemnity agreement contained in the Lease pursuant to DOE's authority in Section 170d of the AEA is sufficient to meet the requirements of Section 193( d) of the Atomic Energy Act of 1954, as amended. Section 193( d) states that "the 1-70

. FOR INFORMATION ONLY License Application for the American Centrifuge Plant Proposed Change 1.4 Application Codes and Standards May 2024 The ACP utilizes a number of the facilities that were originally constructed to support the GCEP and the GDP. The buildings/facilities were designed and constructed according to DOE requirements and/or nationally accepted codes and standards applicable at the time. Many of those codes and standards were earlier versions of current codes and standards that are utilized today for new construction. The codes and standards of record will be verified and documented during the ACP design verification process discussed in Section 11.1.6 of this license application. Any deviations from the codes and standards of record will be evaluated and documented in accordance with the Configuration Management Program as described in Section 11.1 of this license application. New buildings/facilities/processes will meet the codes and standards applicable at the time the facility is designed and constructed as stated in plant design criteria. Modifications to existing buildings and/or facilities will be evaluated to determine if there is a safety benefit from applying current codes and standards and justification will be documented if current codes and standards are not applied.

The following sub-sections list the various industry codes and standards that have been referenced in this license application. The extent to which the Licensee satisfies the requirements of each code or standard is identified individually in the sub-sections. In the context of this section, the terms provisions and guidance are intended to refer only to the explicit requirements of each code or standard.

To establish definitive guidance for the design of the ACP, the Licensee proposed that the license be conditioned as follows:

The Licensee will obtain prior NRC review and approval before deleting or modifying the commitment to any code or standard contained in Section 1.4 of the License Application.

1.4.1 American National Standards Institute/American Nuclear Society ANSI/ ANS 3.1-1987, Selection, Qualification, and Training of Personnel for Nuclear Power Plants The Licensee utilizes the provisions contained in 4.3.3, 4.4.5, and 4.5.3.2 of this standard to develop qualifications of radiation protection personnel.

For the reference to this standard, see Section 4.5.4 of this license application.

ANSI/ ANS 3.2-1994, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants

• 1-119

license Application/or the American Centrifuge Plant Proposed Change May 2024 The Licensee utilizes the provisions contained in Appendix A.6, paragraph (a) of this standard.

For the reference to this standard, see Section 11.4.2.1 of this license application.

ANSVANS 8.1-2014, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors The Licensee satisfies the guidance of this standard with the following exceptions/clarification:

Section 4.1.6 - Operations are reviewed annually; however, personnel in the operating group who are knowledgeable of the NCS requirements for their operations perform this review. Personnel who are knowledgeable in NCS and are independent of operations (e.g., Engineering) provide assistance in these annual reviews.

Personnel who are knowledgeable in NCS and are independent of operations (e.g., Engineering) review operations annually.

For references to this standard, see Sections 5.4.l, 5.4.2, 5.4.5.1, and 5.4.5.2 of this license application and Section 3.10.6 {including subsections) of the ISA Summary for the ACP.

• ANSI/ ANS-8.3-1997, Criticality Accident Alarm System The Licensee satisfies the provision of this standard as modified by Regulatory Guide 3.71 with the following exceptions/clarifications:

Section 1.2.5 - The primary radiation alarm system is the Criticality Accident Alarm System designed to detect a nuclear criticality and provide annunciation using audible alarms that are supplemented by visual alarms in some locations (e.g.,

in high-noise areas) that will alert personnel to evacuate the immediate area. ACP primary facilities that handle 235U in quantities greater than 700_g have Criticality Accident Alarm System coverage e~1cee13t the Uf:& eylinder storage yardsunless the NRC has granted an exemption from the 10 CFR 70.24 CAAS requirements.

For reference to this standard, see Sections 5.4.1, and 5.4.4, and 8.1.1 of this license application; Section 2.2.4 of the Emergency Plan for the American Centrifuge Plant; Qfle-Section 3. 10.6 (including subsections) of the ISA Summary for the ACP: and Section 3.10.4 of Addendum I of the ISA Summary for the ACP -

HA LEU Demonstration.

• ANSI/ ANS-8.19-2014, Administrative Practices for Nuclear Criticality Safety The Licensee satisfies the provisions of this standard with the following exceptions/clarification:

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License Application for the American Centrifuge Plant Proposed Change May2024 The Nuclear Safety Manager can modify the mm1mum qualified NCS Engineer qualification requirements for personnel who have worked for a minimum of three years at other facilities as an NCS Engineer. The Nuclear Safety Manager 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.

For references to this standard, see Sections 5.2.2, 5.4.1, and 11.3.1.8 of this license application.

1.4.2 American National Standards Institute ANSI Nl3.6-1999, Practice for Occupational Radiation Exposure Records Systems The Licensee utilizes the provisions contained in Sections 4, 5, 6, and 7 of this standard for determining radiation protection exposure records.

For the reference to this standard, see Section 4.8.5 of this license application.

ANSI N323-1978, Radiation Protection Instrumentation Test and Calibration The Licensee satisfies the provisions of this standard, except for Sections 4.6 and 5.1 (3).3.

For the reference to this standard, see Section 4.8.4 of this license application.

ANSI Nl4.l-2012, Nuclear Materials - Uranium Hexafluoride - Packagings for Transport The Licensee satisfies the provisions of this standard, except for portions superseded by Federal Regulations with the following exceptions/clarifications:

A. Cylinders, Valves, and Plugs: Cylinders, valves, and plugs are manufactured or purchased to ANSI N14.l-2012._Previously procured and manufactured cylinders, valves, and plugs that meet previous versions of the ANSI standards or specifications in effect at the time of manufacture may be used. Alternatively, existing cylinders, valves, and plugs manufactured to previous version of the ANSI standards or specifications may be modified to meet ANSI N14.l-2012 at some point in the lifecycle due to potential issues or constraints that prohibit continued compliance with standard or specification in effect at the time of manufacture. Only cylinders, valves, and plugs of models still authorized by ANSI N14.l-2012 for manufacture may be accepted for this modification.

Cylinders of this type may be subsequently transferred to the ACP.

B. Cylinder Plugs: Use of steel or aluminum-bronze plugs in UF6 cylinders was acceptable at the United States Enrichment Corporation GDP's for the following operations: heating, feeding, sampling, filling, transferring between cylinders, and 1-122

License Application for the American Centrifuge Plant Proposed Change May 2024 onsite transport and storage. Therefore, these cylinders with these types of plugs may be subsequently transferred to the ACP.

For the reference to this standard, see Section.§. I.J.5.5.5 and 1.2.5 of this license application; Sections 2.2.3 (including subsections), 3.5.5, 3.6.4.1, and 3.7.4 (including subsections) of the ISA Summary for the ACP; and Sections 3.10.4 (including subsections), 7.3.4.4, 7.3.6.4.3. 1,- 7.3.6.7.1.l, &Rd-7.3.6.7.3.J, and Appendix__Q-e of Addendum 1 of the ISA Summary for the ACP - HALEU Demonstration.

1.4.3 American National Standards Institute/American Society of Mechanical Engineers ANSI/ASME NQA-1-2008 and NQA-la-2009 Addenda, Quality Assurance Requirements for Nuclear Facility Applications The Licensee satisfies the provisions of this standard as stated below, with clarification stated in the QAPD:

A. The Licensee satisfies the definitions, as stated in ASME NQA-1-2008 with NQA-1 a-2009 addenda, Part 1, Introduction, Section 400 Terms and Definitions.

8. Indoctrination and training satisfies the provisions of ASME NQA-1-2008, Part l, Requirement 2, Section 200 Indoctrination and Training and Section 500 Records.

C. Personnel performing inspection and testing, as well as QA audit personnel, meet the requirements of ASME NQA-1-2008, Part l, Requirement 2, Section 300 Qualification Requirements and Section 400 Records of Qualification.

D. Design controls that consist of computer programs are developed, validated, and managed in accordance with ASME NQA-1-2008 with the NQA-la-2009 addenda, Part I, Requirement 3, Design Control, Section 800, Requirement 11 Test Control and Part Il, Subpart 2.7 Quality Assurance Requirements for Computer Software for Nuclear Facility Applications.

E. Methods of design verification satisfy the provisions of ASME NQA-1-2008, Part

[, Requirement 3, Section 50 l Methods.

F. Computer Program Testing is performed in accordance with ASME NQA-1-2008 with the NQA-1 a-2009 addenda, Part I, Requirement 11, Test Control.

G. Lifetime records are defined in accordance with ASME NQA-1-2008, Part l, Requirement 17, Section 40 l Lifetime Records.

H. Hard copy or microfilm storage facilities satisfies the guidance of ASME NQA 2008, Part I, Requirement 17, Section 600 Storage.

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NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision 2

2.

Final Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Portsmouth, Ohio Site, DOE/EIS-0360, U. S. Department of Energy Oak Ridge Operations -

Office of Environmental Management, June 2004, Website: http://web.ead.anl.gov/uranium/documents/index.cfm

3.

Form 10-Q, for the quarter ended June 30, 2008

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U.S. Bureau of the Census, 2000, "Population, Housing Units, Area, and Density: 20 IO -

State - Place and (in selected states) County Subdivision 2010 Census Summary File l",

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http:/ /factfi nder.census. gov /bkmk/tab I e/ I. 0/en/D EC/I O _ SF I /GCTPH 1. ST I 0/0400000U S 39

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329-10-002, ACP Memo dated October 15, 2010, Worker and Transient Populations in and around PORTS DOE Reservation, as of October 2010, S. E. Keller

6.

LA-3605-0002, Environmental Report for the American Centrifuge Plant

7.

LA-3605-0003A, Addendum I of the /SA-Integrated Safety Analvsis Summary for the American Centrifuge Plant-HALEU Demonstration

8.

United States National Oceanic and Atmospheric Administration, National Environmental Satellite Data, and Information Service, National Climactic Data Center, Asheville, NC, Climatology of the United States, No. 81, 33 Ohio, Monthly Station Normals of Temperature, Precipitation, and Heating and Cooling Degree Days 1971-2000, February 2002, [NOAA 2003b]

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Huff, Floyd A. and Angel, James R., Rainfall Frequency Atlas of the Midwest, Bulletin 71 (MCC Research Report 92-03) Midwestern Climate Center, Climate Analysis Center, National Weather Service, National Oceanic and Atmospheric Administration, Illinois State Water Survey, A Division of the Illinois Department of Energy and Natural Resources [NOAA 2003c]

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Ohio Department of Natural Resources, Website accessed September 4, 20 I 9, http://parks.ohiodnr.gov/lakewhite I l.

U.S. Department of the Interior, U.S. Geological Survey, Reston, VA, and Website:

http://www.usgs.gov/index.html

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Tetra Tech, Inc. correspondence, Methodology for the 5-mile Population Grids, November 2002 1-143

License Application for the American Centrifuge Plant Proposed Change May 2024

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Construction Materials for Process Gas Applications in Gaseous Diffusion Cascades (U),

GA T-T-3000, Part 8, April 1, 1977

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S. C. Blue and D. E. Underwood, The Corrosion of Highly Alloyed Metals by Fluorinating Gases, KY /L-1990, August I 0, 1990

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Depleted Uranium Hexafluoride Conversion Facility Documented Safety Analysis, DUF6-X-G-DSA-00 1, Revision 2

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Nolen, C. and Rhoden W., Summary of ACP Seismic Design Values, EE-3100-0003, Revision 2, June 2021

30.

Final Report of Subsurface Exploration and Geotechnical Engineering Evaluation, USEC American Centrifuge, Piketon, Ohio, Prepared by Engineering Consulting Services, LLC, ECS Project No. 14-03046, March 2006

31.

Geotechnical Investigation -

American Centrifuge Plant, Project No. F ACP-2063, Prepared by Fugro, William, Lettis and Associates [nc., June 20 I 0

32.

Menne, Matthew J., Imke Durre, Bryant Korzeniewski, Shelley McNeal, Kristy Thomas, Xungang Yin, Steven Anthony, Ron Ray, Russell S. Vose, Byron £.Gleason, and Tamara G. Houston (2012): Global Historical Climatology Network - Daily (GHCN-Daily),

Version

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[USC00338830].

NOAA National Climatic Data Center.

doi:1 0.7289N5D21VHZ, accessed on December 3, 2019

33.

Anthony Arguez, Imke Durre, Scott Applequist, Mike Squires, Russell Vose, Xungang Yin, and Rocky Bilotta (2010). NOAA's U.S. Climate Normals (1981-2010).

[USC00338830].

NOAA National Centers for Environmental Information.

DOI: I0.7289/V5PN93JP, accessed on December 3, 2019

34.

Keller Ohio Office of Strategic

Research, Population Projections, https://development.ohio.gov/files/research/P6090.pdf, accessed on February 5, 2020

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Missouri Census Data

Center, http://mcdc.missouri.edu/cgi-bin/broker? PROGRAM=apps.capsACS.sas& SERVICE=MCDC long& debug=&latit ude=39.0 l 2&1ongitude=83.00 l 4&radii=5&sitename=&dprofile=on&eprofile=on&sprof

36.

37.

38.

i le=on&hprofile=on&units=+&cntypops=on&printdetail=on National Center for Education Statistics, Public School

Data, https://nces.ed.gov/ccd/districtsearch, accessed on February 11, 2020 Kaylor, Keith, Record of Conversations with community facilities, February 11 - 18, 2020 LA-2605-0001, License Application for the American Centrifuge Lead Cascade Facility in Pikeron, Ohie, PiketoA, Ohio 1-145

License Application for the American Centrifuge Plant Proposed Change May 2024

65.

Geraghty & Miller, Analysis of Long-Term Hydrologic Budget for the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, October 1988-September 1989, Dublin, Ohio, 1990

66.

Geraghty & Miller, Quadrant II, RFI Draft Final Report, for the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, Dublin, Ohio, 1992

67.

ERCE, Portsmouth Gaseous Diffusion Plant Final Safety Analysis Report, Section 3.6, Geology and Seismicity, 1990

68.

Ohio Geological Survey, Recent Ohio/ Regional Earthquakes, http://geosurvey.ohiodnr.gov/earthguakes-ohioseis/guakes-felt-in-ohio/recent-ohio-regional-guakes

69.

FBP-ER-RCRA-WD-RPT-0288, Portsmouth Gaseous Diffusion Plant Annual Site Environmental Report-2017

70.

Appendix 1 Lease Agreement between the US. Department of Energy and United States Enrichment Corporation for the Gas Centrifuge Enrichment Plant (GCEP Lease Agreement), as amended

71.

Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores, Revision 3

72.

ASCE 7-2002, Minimum Design Loads for Buildings and Other Structures

73.

K-DA-603, Revision 2, Gas Centrifuge Enrichment Plant General Design Criteria, DOE, February 1982

74.

HALED Demonstration Cascade Completion and HALED Production, Contract Number 89243223CNE000030, as amended

75.

SP-3605-0041, Security Plan for the Protection of Classified Matter at the American Centrifuge Plant

76.

SP-3605-0042, Security Plan for the Physical Protection of Special Nuclear Material at the American Centrifuge Plant 1-148

License Application for the American Centrifuge Plant Proposed Change Blanlc Page 1-149 May 2024

License Application for the American Centrifuge Plant Proposed Change 5.0 NUCLEAR CRITICALITY SAFETY May 2024 The American Centrifuge Plant (ACP) enriches uranium hexafluoride (UF6).

The commercial ACP operation is designed to enrich and safely handle up to 10 weight fwtj percent3/4 uranium-235 (235U). The HA LEU Demonstration Program is designed to enrich and safely handle uranium with an operational limit less than 20-:G weightwt:- percent 235U; however, enrichment levels up to 25 weight wt,-1/4percent 235U 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). The Licensee is required to comply with the performance requirements of 10 Code of Federal Regulations (CFR) 70.6 1. 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 the ACP NCS Program.

In accordance with the requirements contained in l O CFR 70.62, the likelihood and risks of an inadvertent nuclear criticality are evaluated in the Integrated Safety Analysis (ISA). The evaluation considers 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 operation. lf changes to the NCSEs or NCSRs are identified, 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 includes credible nuclear criticality accident scenarios 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 IO CFR 70.6l(d).

The plant has established a threshold of 1 weightwt:- percent or higher enriched 235U and I 00 grams (g) or more of 235U for determining when an evaluation for NCS considerations of planned operations must be performed. This l 00 g 235U mass is a factor of 7 below the minimum critical mass, regardless of whether the material is optimally moderated and fully reflected. Based on this, the value is sufficiently low to use as a threshold limit. In view of this threshold, many of the ACP NCS Program features described in this chapter may not be required to be implemented for operations below the threshold. 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.

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license Application/ or the American Centrifuge Plant Proposed Change 5.2 Organization and Administration 5.2.1 Nuclear Criticality Safety Responsibilities May 2024 The ACP organization and administration are described in Chapter 2.0 of this license application. The Enrichment Operations Plant 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 ofNCS requirements. The Nuclear Safety Manager is responsible for implementing the ACP NCS Program. The management reporting structure for the ACP is depicted in Chapter 2.0 of this license application. The Nuclear Safety Manager has direct access to the Enrichment Operations Plant Manager for nuclear safety matters and reports directly to the Director, Regulatory Affairs.

The ACP organization managers are responsible for ensuring that operations involving uranium enriched to l weight....+- percent or higher 235U and l 00 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 Enrichment Operations Plant 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. l.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).

The fissile material operators are responsible for conducting operations in a safe manner in compliance with procedures and are required to stop operations if unsafe conditions exist.

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

The Nuclear Safety Manager 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 5-3

license Application/or the American Centrifuge Plant Proposed Change May 2024 evaluation to determine reportability (e.g., an event involving the loss of all controls, such that a criticality accident is possible). lf it cannot be determined whether an NCS incident requires reporting under paragraph (a) of l O CFR Part 70, Appendix A, the NCS incident should be reported within one-hour of discovery. Twenty-four-hour reportable events have less safety significance than one-hour reportable events and sometimes require more extensive evaluation to determine reportability. The twenty-four-hour time period for reportable events is intended to allow sufficient time to make this determination. If the determination cannot be completed within this time frame, then the NCS incident is reported within twenty-four hours of discovery. The time of discovery 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 expected to understand that the condition or event adversely impacts double contingency and l O 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.

Records of actions taken are retained in accordance with RMDC requirements described in Section 11.7 of this license application.

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-8.1-2014, ANSI/ANS-8.3-1997, ANSl/ANS-8.19-2014, ANSI/ANS-8.20-1991, ANSl/ANS-8.21-1995, ANSI/ ANS-8.23-2007, 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 Nuclear Criticality Safety Evaluation Each operation involving uranium enriched to I weightwb percent or higher 235U and l 00 g or more of 235U 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.

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license Application/or the American Centrifuge Plant Proposed Change May 2024 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 weight¥.4.- percent or higher 235U and with mass of I 00 g or more 235U (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 contingencies, presenting NCS concerns. Engineeringjudgment 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-8.1-2014, Section 4.2.2, is "Process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible." The phrase "changes in process conditions" does not imply that reliance on two different parameters is required to satisfy 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 conditions would have to occur before a criticality is possible. Process conditions include the characteristics of a process that have an effect on nuclear criticality safety, such as parameters, environment, and operations. Controls are applied 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 procedures, or by other means that limit parameters within specified values. Application of the double contingency 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 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 parameters or conditions relied on and the limits must be specified and justified in the NCSE. Reliance on two different parameters is preferred over reliance on multiple controls on a single parameter. If relying on two or more controls on a single parameter, diverse (i.e., different means of controlling the parameter) is preferred over redundant means of control. Management measures described in Chapter 11.0 of this license application and other safety programs are sometimes used to help ensure a change in a 5-1 0

License Application/or the American Centrifuge Plant Proposed Change May 2024

- After approval by the Nuclear Safety Manager, 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. Editorial changes require only the approval of the Nuclear Safety Manager. 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 operating 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, the conditions delineated in the NCSE. Operational aids such as postings, labels, boundaries for fissile material operations, and fissile material movement guidelines may be used to implement the NCSE. The manager/supervisor ensures postings and labels are prepared and verify that they are properly installed to support implementation of the NCSE. The procedures are prepared or modified to incorporate the NCSE requirements. Managers/supervisors are responsible for ensuring the employees understand the procedures 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 of this 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. If NCS 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.

5.4.2.1 Non-Fissile Material Operations Operations in which the uranium enrichment is less than I weight percent or an inventory of less than 100 g m u are termed **non-fissile material operations" and are performed without the need for NCS double contingency controls. Some opeFatioAs iA'f'Oh'e situatioAs iA whieh the urnRium has aR eRriehmeRt of less thaR I v,ct. pereeRt mu or aR i1weRtory of less thaR 100 g mu. These operatioRs are termed "ROA fissile material operatioRs" aRd are perfofffled without the Reed for NCS double eoRtiRgeRO)' eoRtrols. 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. 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 weight-wt-: percent m u or below 100 g 235U. 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.

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License Application/or the American Centrifuge Plant Proposed Change 5.4.3 Design Philosophy and Review May 2024 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 maintain them if they may credibly degrade such that control of the parameter is jeopardized.

The preferred design approach establishes a 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 (I) passive engineered, (2) active engineered, (3) enhanced administrative, and ( 4) simple administrative controls. The adherence to the preferred design approach is 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/ ANS-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. CAAS is required in each area where special nuclear material is handled. used, or stored unless the NRC has granted an exemption from the IO CFR 70.24 CAAS requirements. In general, coverage is provided for fissile material operations with the following exceptions:, e~,eept the CAAS coverage is not required for UF 6 cylinder storage yards for commercial ACP operations unless the cylinders contain licensed material greater than 5 weightWh percent 235LJ.

CAAS coverage is not required for the handling, storage, and transportation of fissile 30-series UF6 cylinders used during HA LEU Demonstration because there is a trivial risk of criticality for these operations.

as speeified in Reference Section 1.2.5 of this license application for more details. 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 5-13

License Application for the American Centrifuge Plant Proposed Change May 2024 ongoing. Conclusions of non-credibility require at a minimum that the inventory of235U in the area is less than 700 g. ln addition, CAAS is not required for storage areas in which the only special nuclear material present is contained in packages as defined having material that is either 13aekaged or stored in accordance with 10 CFR Part 71 or specifically exempt according to IO CFR 71.15.

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 radiation levels that would result from the minimum criticality accident of concern as defined by ANSI/ ANS:-8.3-1997 and to provide annunciation 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).

For each area requiring CAAS coverage, a monitoring system is installed that provides coverage of the area by at least one detection unit. A detection unit is a set of at least three 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 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 radiation warning lights to deter personnel from entering the area after an evacuation. In addition, facilities within 125 feet of a fissile material operation area requiring CAAS coverage have radiation evacuation horns installed inside and radiation warning lights installed to prompt evacuation and deter personnel 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. 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 regarding nuclear accident planning and response is discussed in Chapter 8 of this license application.

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license Application/or the American Centrifuge Plant Proposed Change 5.4.4.1 Portable CAAS May 2024 In the event a fissi le 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. Alarm annunciation, however, is usually limited to the immediate area within the audible range of the unit's alarm with an additional telemetric link to the X-3012 ACR. 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.

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 125:-foot radius of the fissile material until the operations are concluded. ln 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 application 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 parameter 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 between 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 basis for double contingency, independent means of measurement are used.

• ~ Safety limits on controlled parameters are established and/or implemented with sufficient margin to account for tolerances and uncertainties.

The nuclear parameters which can impact nuclear criticality safety 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 optimally moderated UO2F2/water solutions are more reactive than the oils allowed in the centrifuge process gas equipment or equipment connected to the process gas system. (Reference 16).

When moderation is not controlled either optimum moderation or worst credible moderation is 5-1 5

FOR INFORMATION ONLY License Application/or the American Centrifuge Plant Proposed Change May 2024 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, 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 UF6. This closed system minimizes the introduction of moderation due to wet air in-leakage.

Because UF 6 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). 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 equipment 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.

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license Application/or the American Centrifuge Plant Proposed Change Geometry May 2024 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 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 m u enrichment, composition, etc. Safe mass values may be obtained from established standards or operation specific reactivity calculations. "Safe mass" is defined as the quantity of fissi le material that is safely subcritical under the most reactive credible conditions (defined for a given isotopic composition and physiochemical form), including allowance for over-batching. 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. Unless specifically controlled, an item containing enriched uranium is assumed to contain the most 235U credible based on the available volume. When mass is determined through measurement, instrumentation that is subject to management measures is used.

Enrichment The maximum 235U enrichment for each operation is established by the specific NCSE.

Credible process upset conditions that could alter the mu enrichment are also considered in the NCSEs.

When the enrichment of uranium needs to be measured for an NCS control, the measurement is obtained using either installed equipment or based on samples analyzed m a laboratory.

Uranium-containing material in the ACP with 235U enrichment less than 1 weight-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 I weight-wt:- percent.

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License Application/or the American Centrifuge Plant Proposed Change May 2024 The 235U enrichment ofUF 6 in the ACP HALEU cascade is -limited to less than 20 weightwt-:-

percent with the potential for momentary enrichment transients up to 25 weight~ 1/4-percent 235U during HALEU cascade operations. Small quantities of greater than 20 weight~ percent 235U may also be present outside of plant equipment in the form of standards.

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.

UF 6 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.

Heterogeneity Heterogeneous configurations are considered for those operations 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 of heterogeneity are based on observed physical characteristics.

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 samples are independently analyzed as required by the specific NCSE for the operation involved. The specific controls and details are documented in the NCSE for each operation that relies on concentration controls. Precipitating agents, including freezing, are controlled as necessary to ensure they do not inadvertently affect solubility or homogeneity or increase the concentration.

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

The possibility of full water reflection is considered when performing analyses. 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 fu ll density water blocks to simulate personnel. It is recognized that concrete can be a more efficient reflector than water, and its potential presence is considered. lf special moderators such as deuterium, beryllium, or graphite, or iflarge amounts of hydrogen-rich materials (e.g., hydrocarbon oil or polyethylene, etc.) are present, the NCS evaluation ensures the modeled reflection conditions remain bounding. Reflection controls are used to limit the potential reactivity of a fissile material operation.

5-18

License Application for the American Centrifuge Plant Proposed Change 5.5 References May 2024

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

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

3. ANSI/ ANS-8.19-2014, Administrative Practices for Nuclear Criticality Safety

4. ANSI/ANS-8.20-1991, Nuclear Criticality Safety Training

5. ANSI/ANS-8.21-1995, Use ofFixedNeutronAbsorbers inNuclear Facilities Outside Reactors

6. ANSI/ANS-8.23-2007, Nuclear Criticality Accident Emergency Planning and Response

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

8. ANSI/ANS-8.26-2007, Criticality Safety Engineer Training and Qualification Program

9. ARH-600, Criticality Handbook, Volumes I, II, and III, Atlantic Richfield Hanford Co. Report (1968)

10. LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant

11. LA-10860-MS, Criticality Dimensions of Systems Containing 235U, 239Pu, and 233U, 1986 Revision

12. NRC Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Material Facilities, Revision 3

13. NUREG-1513, Integrated Safety Analysis Guidance Document

14. NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision2

15. EE-3101-0013, NCS Code Validation of SCALE 6.2.3 and Cross Section Set v7-252 for keff Calculations, Rev. 0, December 2019

16. DAC-3101-0006, Safe Mass Study for UF4 and Oil, February 2020

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

NEA/NSC/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 ofKENO V.a Comparison with Critical Experiments," ORNL/CSD/TM-238, Martin Marietta Energy Systems, Contract Number DE-AC05-84OR21400, December 1986-;-

5-22

License Application for the American Centrifuge Plant Proposed Change 8.0 EMERGENCY MANAGEMENT May 2024 As discussed in Section 1.1.8 of this license application, it is the long-term goal of the Licensee to deploy the American Centrifuge Plant (ACP) in a modular fashion on a scalable, economical gradation consistent with market demand. An;ierican Centrifuge Operating, LLC (ACO), the Licensee, would develop and submit future license amendments to allow additional phases of modular deployment up to the currently U.S. Nuclear Regulatory Commission (NRC)-

approved full capacity operation of 3.8 million separative work units. Pursuant to 10 Code of Federal Regulations (CFR) 70.22(i), the Licensee developed an NRC-approved Emergency Plan for the fully deployed -ACP and other on-going activities on the U.S. Department of Energy (DOE) reservation in Pike County Ohio. The previously NRC-approved plan conforms to the Regulatory Guide 3.67, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities, dated January 1992. Although not required or implemented for the High Assay Low Enriched Uranium (HALED) Demonstration Program, the Emergency Plan will support future ACP deployment phases.

The information documented in the previously NRC-approved emergency plan includes:

1) description of the facility; 2) summary credible emergencies; 3) classification and notification of accidents; 4) responsibilities; 5) emergency response measures; 6) equipment and facilities designated for use during emergencies; 7) methods for maintaining emergency preparedness; 8) emergency records and reports; 9) recovery and restoration measures; and 10) a commitment to comply with the Community Right-To-Know Act.

The previously NRC-approved plan remains as part of this license application as document NR-3605-0008, Emergency Plan for the American Centrifuge Plant in Piketon, Ohio.

The Licensee would notify the NRC well in advance of the transition into any future phases of deployment that would require use of this previously NRC-approved NR-3605-0008.

8.1 High Assay Low Enriched Uranium Demonstration No Emergency Plan as discussed under 10 CFR 70.22(i) is needed for the HALED Demonstration Program. DAC-3901-0005, Evaluation of No Need for an Emergency Plan for the HALEU Demonstration, provides the evaluation stipulated in 10 CFR 70.22(i)(l)(i) to demonstrate that no Emergency Plan is required for the HALED Demonstration Program. The evaluation shows that the maximum dose to a member of the public offsite due to a release of radioactive materials would not exceed 1 roentgen equivalent man (rem) effective dose equivalent or an intake of 2 milligrams (mg) of soluble uranium (U).

Fluor BWXT Portsmouth, LLC (FBP); Portsmouth Mission Alliance, LLC; and Mid-America Conversion Services, LLC are the DOE's primary contractors at the DOE Portsmouth site.

FBP currently serves as the Decontamination and Decommissioning contractor and provides emergency response capabilities at the site compliant under DOE Order 151.lD, Comprehensive Emergency Management System.

Through a reverse work authorization arrangement, FBP provides emergency response to the ACP.

With augmentation and 8-1

License Application for the American Centrifuge Plant Proposed Change coordiation with ACO personnel where appropriate, FBP provides the following:

Emergency Response Organization Emergency Facilities and Equipment/Systems Fire Department Response*

Emergency Operations Center (EOC)

Alternate EOC Joint Information Center Plant Shift Superintendent/Incident Command Support Emergency Medical Support Offsite Response Interfaces Announcement of Protective Actions Emergency Public Information Communications and Notifications, as appropriate Consequence Assessment Support with Termination and Recovery, as appropriate Support with Coordinating and Assessing Readiness Assurance May 2024

• FBP operates and maintains the X-1007 Fire Station on the DOE reservation. This is a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day/7 days a week dedicated fire department which has minimum staffing requirements to maintain appropriate manpower for emergency response on the DOE reservation. Fire personnel are certified as State of Ohio Level II (Professional Firefighters) and minimum State of Ohio Emergency Medical Technicians with one Paramedic per shift.

8.1.1 Nuclear Criticality The primary radiation alarm system is the Criticality Accident Alarm System (CAAS),

designed to detect a nuclear criticality and provide annunciation by audible evacuation alarms that are supplemented by visual alarms in some areas, such as high-noise areas that will alert personnel to evacuate the immediate area.

Operations involving fissile material are evaluated for Nuclear Criticality Safety (NCS) considerations prior to initiation.

The need for CAAS coverage is considered during the evaluation process. CAAS coverage is required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from the 10 CFR 70.24 CAAS requirements. Exemptions from the requirements of IO CFR 70.24 are documented in Section 1.2.5 of this license application. CAAS eoverage is 19ro*rided, unless it is determined that eo*,*erage is not required 19er the requirements of IO CFR 70.24 and the finding is doeumented in an l'JC8 g.,,aluation. CAAS coverage is provided for HALEU Demonstration fissi le material operations.

The CAAS is designed to detect gamma radiation levels that would result from the minimum criticality accident of concern as defined in I0_CFR_70.24(a)(l). The CAAS is designed to provide annunciation by audible alarms that are supplemented by visual alarms in some areas, such as in high-noise areas.

8-2

License Application/or the American Centrifuge Plant Proposed Change May 2024 The criticality detection system consists of detector clusters and an alarm system. When a criticality accident alarm activates, a radiation alarm is generated actuating building local horns. Alarm activation requires evacuation of personnel from the affected area to a designated monitoring station that is located a minimum evacuation distance of 125 ft from the facility with the active CAAS alarm. Trained emergency responders are dispatched to the facility evacuation point to provide evacuees and Incident Command with additional guidance, as appropriate.

Based on the alarm location, Incident Command can direct the actions necessary to respond to the accident in coordination with technical personnel. Emergency response to CAAS alarms and/or nuclear criticality events is consistent with guidance contained in ANSI/ANS-8.23-2007, Nuclear Criticality Accident Emergency Planning and Response.

Coordinated response exercises and local drills are performed periodically to familiarize personnel with proper response actions and assembly locations.

8.2 References I. Reference DeletedAmerican *National Standards Institute (ANSl)/Ameriean ~~uelear Society (A~JS) 8._3 1997, C,,itieality Aeeident AJ.a,.,'ji 8y'Sff!/H

2. American National Standards Institute (ANSI)/American Nuclear Society (ANS) 8.23-2007, Nuclear Criticality Accident Emergency Planning and Response

3. Regulatory Guide 3.67, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities, Revision 1

4. NR-3605-0008, Emergency Plan for the American Centrifuge Plant

5. DAC-3901-0005, Evaluation of No Need for an Emergency Plan for the HALEU Demonstration

6. NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision 2

7. DOE Order 151.1 D, Comprehensive Emergency Management System 8-3 of ACO 24-0035 Proposed Changes to NR-3605-0008, Emergency Plan for the American Centrifuge Plant Information Contained Within Does Not Contain Export Controlled Information Reviewing Official:

Lori Hawk/ ACO Date:

05/01/2024

NR-3605-0008 Emergency Plan for the American Centrifuge Plant in Piketon, Ohio Docket No. 70-7004 The American CentrifugeT" Proposed Change Information contained within does not contain Export Controlled Information Reviewer: Lori Hawk / ACO Date: 05/0112024 May 2024

Emergency Plan/or the American Centrifuge Plant Proposed Change 2.2.3 Other Toxic Chemical Releases May 2024 Detection equipment and/or chemical release alarms for various toxic chemicals in the plant have been installed at strategic locations where particular chemicals are present. As in a Uf 6 release, if an operator is in the immediate vicinity of a chemical release, the operator should detect the release by sight or smell. Upon recognition or detection of a release, the release is reported immediately to fire protection personnel and the PSS.

Both the fire protection personnel and PSS (acting in the role of the IC) respond to the incident area upon receiving an indication of a chemical release.

2.2.4 Nuclear Criticality The primary radiation alarm system is the CAAS, designed to detect a nuclear criticality and provide audible and visual alarms that will alert personnel to evacuate the immediate area.

Operations involving fissile material are evaluated for NCS considerations prior to initiation. The need for CAAS coverage is considered during the evaluation process. CAAS coverage is pro;rided, uAless it is determiAed that coverage is AOt required in each area where special nuclear material is handled, used, or stored unless the NRC has granted an exemption from ~

the requiremeRts of 10 CFR 70.24 CAAS requirements. aAd the fiAdiAg is documeAted iA aA *NcS b'l1aluatioA. Exemptions from the requirements of IO CFR 70.24 are documented in Section 1.2.5 of the license applicationCAAS co;rerage is pFO'l'ided for ACP fissile material operatioAs, e>feept the ui;:e eyliRder storage yards as specified iA SeetioA 1.2.5 of this lieeAse applicatioA.

The CAAS is designed to detect neutron radiation levels that would result from the minimum criticality accident of concern as defined by American National Standards Institute (ANSl)/American Nuclear Society (ANS):-8.3:-fl 997 l:;ditioA for the ACP aAd 1986 edition for the GDP) and to provide an audible evacuation alarm.

The criticality detection system consists of locator clusters and an alarm system. When a criticality accident alarm activates, a radiation alarm is generated actuating building local horns.

Alarm activation requires evacuation of personnel from the affected area to a designated monitoring station that is located a safe distance from the area. On the basis of the alarm location, the IC can direct the actions necessary to respond to the accident. Emergency response to CAAS alarms and/or nuclear criticality events is consistent with guidance contained in ANSI/ ANS-:_8.23-2007 +99+.

2.2.5 Natural Phenomena 2.2.5.1 Earthquake Digital strong motion accelerographs are installed for detecting earthquake-type movements. The strong motion accelerograph units are electronically connected in such a way that if one is triggered, the accelerograph units will start recording. Activation of the seismic 16

Emergency Plan/or the American Centrifuge Plant Proposed Change May 2024

11.0 REFERENCES

I.

2.

3.

4.

5.

6.

7.

8.

9.

11.

12.

13.

14.

U.S. Nuclear Regulatory Commission, NUREG-1 520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility U.S. Nuclear Regulatory Commission, Regulatory Guide 3.67, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities USEC-02, Portsmouth Gaseous Diffusion Plant Application for United States Nuclear Regulatory Commission Certification, Safety Analysis Report, Volumes 1 and 2 USEC-02, Portsmouth Gaseous Diffusion Plant (PORTS) Emergency Plan LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant U.S. Environmental Protection Agency, 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents American Industrial Hygiene Association, Emergency Response Planning Guides (ERPGs)

U.S. Environmental Protection Agency, 40 CFR Part 68, Appendix A, Table of Toxic Endpoints American National Standards Institute/ American Nuclear Society 8.3-1997, l'htelee1r Criticality Accident Alarm SystemEme,"gcney1 Ple1,'ining e1196' Respense, 1997 Reference DeletedAmericaA ~latioRal 8taAdards IAstitllte/A1:nericaA ~lltclear 8ociet)' 8.3 1986, Critice1lity Accide}'lt Ale11w'l Sj,1stcm, 1986 American National Standards Institute/American Nuclear Society 8.23--4-99+2007, Nuclear Criticality Accident Emergency Planning and Response~

U.S. Department of Energy Oak Ridge Operations -

Office of Environmental Management, Final Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Portsmouth, Ohio Site, DOE/EIS-0360, June 2004 Lawrence Livermore National Laboratory, The Engineering Analysis Report for the Long-Term Management of Depleted Uranium Hexafluoride, Volumes 1 and 2, Revision 2, Depleted Uranium Hexafluoride Management Program, UCRL-AR-124080, May 1997 Depleted Uranium Hexafluoride Conversion Facility Documented Safety Analysis, DUF6-X-G-DSA-001, Revision 2 63 of ACO 24-0035 Affidavit Information Contained Within Does Not Contain Export Controlled Information Reviewing Official:

Lori Hawk, ACO Date:

05/09/2024

AFFIDAVIT OF LARRY B. CUTLIP SUPPORTING APPLICATION TO WITHHOLD FROM PUBLIC DISCLOSURE CERTAIN INFORMATION PROVIDED TO NRC IN LETTER ACO 24-0035 I, Larry B. Cutlip, of American Centrifuge Operating, LLC (ACO), having been duly sworn, do herby affirm and state:

1. I am the President of ACO and have been authorized by ACO to (a) review the information owned by ACO which is referenced herein and attached hereto relating to a license amendment request to provide and clarify exemptions from the requirements of 10 CFR 70.24, Criticality

_Accident Requirements, described in letter ACO 24-0035, 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.l 7(a)(4), and (b) apply for the withholding of such information from public disclosure by the U.S. Nuclear Regulatory CommissiQn (NRC) on behalf of ACO, and ( c) sign and file with the NRC this affidavit and the attachments hereto.

1. 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 b~ 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. ACQ has a rational basis for determining the types of information customarily held in confidence by,it and, in that connection, util_izes 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 competitive 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.

1 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 the Commission in confidence and, under the provisions of 10 CPR 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 available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

The proprietary information sought to be withheld is contained within Enclosures 4 and 5 of letter ACO 24-0035. Enclosure 4 provides proposed changes to LA-3605-0003, Integrated Safety Analysis Summary for the American Centrifuge Plant. Enclosure 5 provides proposed changes to LA-3605-0003A, Addendum I of the Integrated Safety Analysis Summary for the American Centrifuge Plant - HALEU Demonstration. These enclosures discuss the types of accidents associated with the American Centrifuge Plant and the HALEU Demonstration Project; therefore, determined to be proprietary.

Public disclosure of this proprietary information is likely to cause substantial harm 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 Security-Related Information and/or Export Controlled Information.

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 and resources would have to be expended.

Further the deponent sayeth not.

On this 9th day of May 2024, 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 he executed the same for the purposes therein contained.

In witness hereof I hereunto set my hand and official seal.

Kathy~r State of Tennessee Notary Public Anderson County My commission expires October 26, 2024