ML23145A248

Louisiana Energy Services, LLC, Safety Analysis Report, Revision 50

ML23145A248
Person / Time
Site:	07003103
Issue date:	05/25/2023
From:	Louisiana Energy Services, URENCO USA
To:	Office of Nuclear Material Safety and Safeguards
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ML23145A246	List: ML23145A247 ML23145A248
References
LES-23-065-NRC
Download: ML23145A248 (1)

Text

SAFETY ANALYSIS REPORT Revision 50

Table of Contents Summary of Changes Rev 50 Issue /

Change Description of Change Date GCP-073:

SAR Section 1.1.3.2, pg 1.1-7, 1.1 Removed references to various Production Phases of Operations in Chapter 12, AC 171129: Updated SAR Section 1.3.2.3, pg 1.3-4 with Covenant Health Hobbs Hospital name change and updated the bed count from 250 to 99 AC 168390 - Updated SAR Section 1.2.1.2, pg 1.2-11 to state "The President and Chief Executive Officer" for consistency Corrected minor issues with SAR 5.6-7, pgs 5.6-7/10/11 tables (blue 50 LBDCR-22-016 text and have inconsistent fonts)

AC 170888 - Updated SAR Section 8, pg 8.0-1 to indicate that the MOU files are kept by the Safety and Emergency Response Manager.

AC 170838: SAR Section 11.5.1, pg 11.5-1 Assessment, moved the nuclear criticality safety requirements from the 3rd paragraph that are applicable to the NCS assessment, into paragraph 7 discussion about the NCS assessment.

AC 169116: Revised table in SAR Section 12, p 12.0-2 to correct the accident sequence, as CHEM RELEASE-WORKER EVAC is not an EE sequence Safety Analysis Report Page-i Rev 50

1.1 Facility and Process Description composition of isotopes 234U, 235U, and 238U or depleted 235U content (i.e., tails). The enrichment process is a mechanical separation of isotopes using a fast rotating cylinder (centrifuge) based on a difference in centrifugal forces due to differences in molecular weight of the uranic isotopes. No chemical changes or nuclear reactions take place. The feed, product, and tails streams are all in the form of UF6.

1.1.3.2 Process System Descriptions An overview of the four enrichment process systems and the two enrichment support systems is discussed below.

Numerous substances associated with the enrichment process could pose hazards if they were released into the environment. Chapter 6, Chemical Process Safety, contains a discussion of the criteria and identification of the chemicals of concern at the NEF and concludes that uranium hexafluoride (UF6) is the only chemical of concern that will be used at the facility. Chapter 6, Chemical Process Safety, also identifies the locations where UF6 is stored or used in the facility and includes a detailed discussion and description of the hazardous characteristics of UF6 as well as a detailed listing of other chemicals that are in use at the facility.

The enrichment process is comprised of the following major systems:

UF6 Feed System LBDCR-22-016 (See 12.2.2.1) The first step in the process is the receipt of the feed cylinders and preparation to feed the UF6 through the enrichment process.

Natural UF6 feed is received at the NEF in 48Ycylinders from a conversion plant. Pressure in the feed cylinders is below atmospheric (vacuum) and the UF6 is in solid form.

The function of the UF6 Feed System is to provide a continuous supply of gaseous UF6 from the feed cylinders to the cascades.

A Solid Feed Station and Feed Purification Low Temperature Take-off Station have the ability to transfer Natural UF6 feed from a 48Y cylinder directly to a Product Low Temperature Take-off Station 30B cylinder, bypassing the cascade system. This is accomplished through a connection made from test valve terminals on either system.

Cascade System LBDCR-22-016 (See 12.2.2.2) The function of the Cascade System is to receive gaseous UF6 from the UF6 Feed System and enrich the UF6 up to the LES license limit in isotope 235U.

Multiple gas centrifuges make up arrays called cascades. The cascades separate gaseous UF6 feed with a uranium isotopic concentration (0.711 w/o 235U or less) into two process flow streams

- product and tails. The tails stream is UF6 that has been depleted of 235U isotope to 0.1 to 0.5 w/ 235U.

o Safety Analysis Report Page-1.1-7 Rev 50

1.1 Facility and Process Description Product Take-off System LBDCR-22-016 (See 12.2.2.3) The function of the Product Take-off System is to provide continuous withdrawal of the enriched gaseous UF6 product from the cascades and to purge and dispose of light gas impurities from the enrichment process.

The product streams leaving the cascades are brought together into one common manifold from the Cascade Hall. The product stream is transported via a train of vacuum pumps to Product LTTS in the UF6 Handling Area. There are five Product LTTS per Cascade Hall.

The Product Take-off System also contains a system to purge light gases (typically air and HF) from the enrichment process. This system consists of UF6 Cold Traps which capture UF6 while leaving the light gas in a gaseous state. The cold trap is followed by product vent Vacuum Pump/Trap Sets, each consisting of a carbon trap, an alumina trap, and a vacuum pump. The carbon trap removes small traces of UF6 and the alumina trap removes any HF from the product gas.

Tails Take-off System LBDCR-22-016 (See 12.2.2.4) The primary function of the Tails Take-off System is to provide continuous withdrawal of the gaseous UF6 tails from the cascades. A secondary function of this system is to provide a means for removal of UF6 from the centrifuge cascades under abnormal conditions.

The tails stream exits each Cascade Hall via a primary header, goes through a pumping train, and then to Tails LTTS in the UF6 Handling Area. There are eight Tails LTTS per Cascade Hall.

In addition to the four primary systems listed above, there are two major support systems:

Product Blending System LBDCR-22-016 (See 12.2.2.5) The primary function of the Product Blending System is to provide a means to fill 30B cylinders with UF6 at a specific enrichment of 235U to meet customer requirements. This is accomplished by blending (mixing) UF6 at two different enrichment levels to one specific enrichment level. The system can also be used to transfer product from a 30B cylinder to another 30B cylinder without blending.

The Product Take-off System also provides a method for transferring natural feed from a 48Y cylinder to a 30B cylinder to support off-site operations. This is accomplished by a connection from a Feed System test valve terminal point to a test valve terminal point leading to a Product LTTS. This method bypasses the cascade system.

This system consists of Blending Donor Stations (which are similar to the Solid Feed Stations) and Blending Receiver Stations (which are similar to the Product LTTS) described under the primary systems.

Product Liquid Sampling System LBDCR-22-016 (See 12.2.2.6) The function of the Product Liquid Sampling System is to obtain an assay sample from filled 30B cylinders. The sample is used to validate the exact enrichment level of UF6 in the filled 30B cylinders before the cylinders are sent to the fuel processor.

Safety Analysis Report Page-1.1-8 Rev 50

1.2 Institutional Information Institutional Information This section addresses the details of the applicants corporate identity and location, applicant's ownership organization and financial information, type, quarterly, and form of licensed material to be used at the facility, and the type(s) of license(s) being applied for.

Corporate Identity 1.2.1.1 Licensee The Licensees name, address, and principal office are as follows:

Louisiana Energy Services, L.L.C.

P.O. Box 1789 275 Highway 176 Eunice, NM 88231 1.2.1.2 Organization and Management of Applicant Louisiana Energy Services (LES), L.L.C. is a Delaware limited liability company. It has been formed solely to provide uranium enrichment services for commercial nuclear power plants.

LES has one, 100% owned subsidiary, operating as a limited liability company, formed for the purpose of purchasing Industrial Revenue Bonds and no divisions. The ownership of LES is as follows:

1. URENCO Investments, Inc. (UII) (a Delaware corporation and wholly-owned subsidiary of URENCO Limited, a corporation formed under the laws of the of England (URENCO) and owned in equal shares by Enrichment Investments Limited (EIL),

Uranit UK Limited (Uranit), both companies formed under English law, and Ultra-Centrifuge Nederland NV (UCN), a company formed under Dutch law. EIL is ultimately wholly-owned by the government of the United Kingdom; UCN is ultimately wholly-owned by the government of the Netherlands; Uranit is ultimately owned by Eon Kenkraft GmbH (50%) and RWE Power Ag (50%), companies formed under the laws of the Federal Republic of Germany). UII holds 96% (as of December 31, 2010) of the membership units and has 100% of the voting power.

2. URENCO Deelnemingen B.V. (a Netherlands corporation and wholly-owned subsidiary of URENCO USA Inc. The ownership of URENCO USA Inc. is explicitly described above. URENCO Deelnemingen B.V. holds 4% of the membership units (as of December 31, 2010) and has 0% of the voting power. LBDCR-22-016 The President and Chief Executive Officer of LES is Karen Fili. The President and Chief Executive Officer reports to the Board of Managers. The Board of Managers are:

Ms. Karen D Fili - Chairperson, Manager Mr. Chris Chater - Manager Mr. Paul Lorskulsint - Manager Mr. David E Sexton - Manager Safety Analysis Report Page-1.2-11 Rev 50

1.3 Site Description 1.3.2.3 Proximity to Public Facilities - Schools, Hospitals, Parks The Eunice First Assembly of God Church is located about 9 km (5.4 mi) from the site.

LBDCR-22-016 There are two hospitals in the vicinity of the site. The Covenant Health Hobbs HospitalLea Regional Medical Center is located in Hobbs, New Mexico about 32 km (20 mi) north of the NEF site. This 25099-bed hospital can handle acute and stable chronic care patients. In Lovington, New Mexico, 64 km (39 mi) north-northwest of the site, Covenant Medical Systems manages Nor-Lea Hospital, a full-service, 27-bed facility.

Eunice Senior Center is located about 9 km (5.4 mi) from the site.

There are four educational facilities within about 8 km (5 mi) of the NEF site, all in Eunice, New Mexico. These include an elementary school, a middle school, a high school, and a private K-12 school.

Eunice Fire and Rescue and the Eunice Police Department are located approximately 8 km (5 mi) from the site.

The Eunice Golf Course is located approximately 14.7 km (9.4 mi) from the site.

1.3.2.4 Nearby Industrial Facilities (Includes Nuclear Facilities)

Nuclear Facilities There are no nuclear production facilities located within 32 km (20 mi) of the site, therefore neither environmental nor emergency preparedness interactions between facilities is required.

Non-Nuclear Facilities The site is bordered to the north by railroad tracks beyond which is a quarry operated by Wallach Concrete Company. The quarry owner leases land space to Sundance Services, a reclamation company that maintains three small produced water lagoons.

Lea County operates a landfill on the south side of Section 33 across New Mexico State Highway 234, approximately 1 km (0.6 mi) from the center of the site.

A vacant parcel of land is immediately east of the site. Land further east, in Texas, is occupied by WCS. WCS possesses a radioactive materials license from Texas, an NRC Agreement state, and is licensed to treat and temporarily store low-level radioactive waste.

Dynegys Midstream Services Plant is located 6 km (4 mi) from the site. This facility is engaged in the gathering and processing of natural gas for the subsequent fractionation, storage, and transportation of natural gas liquids.

An underground CO2 pipeline, running southeast-northwest, originally traversed the property.

This underground CO2 pipeline has been relocated to the western edge of the property boundary.

An underground natural gas pipeline is located along the south property line, paralleling New Mexico Highway 234.

Safety Analysis Report Page-1.3-4 Rev 50

5.6 Chapter 5 Tables LEU-SOL-THERM-005 Boron Carbide A large number of critical experiments with absorber Absorber Rods in elements of different types in uranium nitrate solution Uranium (5.64% 235U) of different enrichments and concentrations were Nitrate Solution performed in 1961 - 1963 at the Solution Physical Facility of the Institute of Physics and Power Engineering (IPPE), Obninsk, Russia. The purpose of these experiments was to determine the effects of enrichment, concentration, geometry, neutron reflection, and type, diameter, number, and arrangement of absorber rods on the critical mass of light-water-moderated homogenous uranyl nitrate solutions. The experiments included ones with a central boron carbide or cadmium rod, clusters of boron carbide rods, and triangular lattices of boron carbide rods in cylindrical tanks of different dimensions filled with solutions of uranyl nitrate.

The three experiments included in this evaluation were performed with uranium enriched to 5.64 w 235

/o U. Uranium nitrate solution with uranium concentration of 400.2 g/l was pumped into the core or inner tank, a stainless steel cylindrical tank with an inner diameter 110 cm, on experiment was performed without absorber rods, another one with a central rod, and another one with a cluster of seven absorber rods arranged at the corners and center of a hexagon with a pitch of 31.8 cm, inserted in the center of the core tank. There was a thick side and bottom reflector in these experiments.

LEU-SOL-THERM-006 Boron Carbide A large number of critical experiments with absorber Absorber Rods in elements of different types in uranium nitrate solution Uranium (10% U-235) of different enrichments and concentrations were Nitrate Solution performed at the Solution Physical Facility of IPPE.

The purpose of these experiments was to determine the effects of enrichment, concentration, geometry, neutron reflection, and type, diameter, number, and arrangement of absorber rods on the critical mass of light-water-moderated homogeneous uranyl nitrate solutions.

The five experiments included in this evaluation were performed with uranium enriched to 10 wt% U-235.

Uranium nitrate solution with uranium concentration of 420.5 g/l was pumped into the core or inner tank, a stainless steel cylindrical tank with inner diameter 110 cm. One experiment was performed without absorber rods. In each of four experiments a different number of boron carbide absorber rods was inserted in the core tank. The absorber rods were arranged in a hexagonal lattice with different pitches. There was a thick side and bottom water reflector in these experiments.

Safety Analysis Report Page-5.6-7 Rev 50

5.6 Chapter 5 Tables IEU-COMP-MIXED-003 Unreflected UF4-CF2 A series of fourteen critical experiments using an (IEU-COMP-INTER-003) Blocks with 37.5% U- equimolar mixture of uranium tetrafluoride (UF4) and 235 Teflon (CF2) were performed at the Oak Ridge National Laboratory (ORNL) Critical Experiments Facility in 1956 and 1957. The uranium was enriched to approximately 37.5 wt% in U-235. The mixture was compressed into three different sizes of blocks to enable stacking of critical configurations of various shapes. Very little moisture was present in this material, so experiments at very low H/X values were possible. In some cases, interstitial moderator material was used to alter the H/X of the overall experiment. All but one of these contained methacrylate plastic, and the remaining one used cellulose acetate plastic. H/X values for the unreflected series of experiments range from 0.07 to 17.10.

IEU-SOL-THERM-002 Bare and Water- A series of experiments a spherical core vessel reflected Spheres and featuring uranium/hydrogen systems at ~30% U-235, Hemispheres of were performed at Dounreay. The experiments Aqueous Uranyl considered five aluminum spheres of various Fluoride Solutions diameters to determine the solution concentration (30.45% U-235) necessary for criticality when a sphere was completely filled.

IEU-SOL-THERM-004 Water Boiler The experiment performed at Los Alamos National Experiment: Beryllium Laboratory involved 14.7% enriched uranyl sulfate Oxide-Reflected (UO2SO4) solution in a 1 ft. diameter stainless steel Sphere Containing sphere with a H/U-235 ratio of 646.

Uranyl (14.7) Sulfate Solution IEU-SOL-THERM-005 Unreflected Critical Several series of critical experiments involving Dimension of Aqueous aqueous uranyl fluoride (UO2F2) solutions were Solution of U (37 wt%) performed at ORNL between the years 1958 and O2F2 in Spherical 1960. These experiments were performed to Geometry determine the conditions under which aqueous solutions of intermediate enriched uranium (37 wt%

U-235) can be made critical. The critical experiments were performed in different geometries (spherical and cylindrical) and at different U concentrations and solution heights in cylindrical geometry. This critical experiment was performed with U(37 wt%)O2F2 in spherical geometry with uranium concentration of 49.15 mg uranium/g of solution.

IEU-MET-FAST-003 Bare Spherical Criticality measurements of bare metal U-235(36 Assembly of U-235 (36 wt%) assemblies were conducted at VNIIEFs wt%) criticality test facility (CTF) in 1994. As a result of these efforts, two benchmark models, one detailed and one simplified, of a single experiment were developed. (The simplified model is evaluated herein.)

Safety Analysis Report Page-5.6-10 Rev 50

5.6 Chapter 5 Tables IEU-MET-FAST-005 Steel-Reflected Criticality measurements of steel-metal U-235(36 Spherical Assembly of wt%) assemblies were conducted at VNIIEFs CTF in U-235 (36 wt%) 1994. As a result of these efforts, two benchmark models, one detailed and one simplified, of a single experiment were developed. (The simplified model is evaluated herein.)

IEU-MET-FAST-009 Spherical Assembly of Criticality measurements of a polyethylenereflected U-235 (36 wt%) with a assembly of U-235 (36 wt%) were conducted by 5.75 cm Polyethylene VNIIEF in 1977 at its criticality test facility. These Reflector efforts resulted in an acceptable spherical benchmark model of the critical assembly.

IEU-MET-FAST-015 ZPR-3 Assembly 6F: A Over a period of 30 years, more than a hundred Zero Spherical Assembly of Power Reactor (ZPR) critical assemblies were Highly Enriched constructed at Argonne National Laboratory. The Uranium, Depleted ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were Uranium, Aluminum all fast critical assembly facilities. The ZPR critical and Steel with an assemblies were constructed to support fast reactor Average U-235 development, but data from some of these Enrichment of 47 atom assemblies are also well suited for nuclear data

% validation and to form the basis for criticality safety benchmarks. ZPR-3 Assembly 6F (ZPR-3/6F) simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core U-235 enrichment of approximately 47 wt%.

IEU-MET-FAST-019 45.5% U-235 Pseudo- A series of sub-critical high-multiplication experiments Cylindrical Metal were performed at the Atomic Weapons Slabs: Bare Establishment (Aldermaston) to determine the critical Assemblies thickness of IEU (45.5 wt% U-235) metal slabs.

Approximately 470 kg of 45.5 wt% enriched uranium metal were available for this work. This material was cast and machined into hexagonal and half-hexagonal blocks of varying thickness. These blocks were then used to construct pseudo-cylindrical assemblies. (The two bare assemblies are evaluated herein).

Safety Analysis Report Page-5.6-11 Rev 50

8.0 Emergency Management Emergency Management The plans for coping with emergencies at the National Enrichment Facility are presented in the facility Emergency Plan. The Emergency Plan has been developed in accordance with 10 CFR 70.22(i) (CFR, 2003a) and 10 CFR 40.31(j) (CFR, 2003b). The Emergency Plan conforms to the guidance presented in Regulatory Guide 3.67, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities. The facility Emergency Plan also addresses the specific acceptance criteria in NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, Chapter 8, Emergency Management.

LBDCR-22-016 The Emergency Plan identifies the offsite organizations that reviewed the Emergency Plan pursuant to the requirement in 10 CFR 70.22(i)(4) (CFR, 2003a) and 10 CFR 40.31(j)(4) (CFR, 2003b). Memorandums of Understanding with the off-site organizations are kept by the Safety and Emergency Response Manager.are provided in the Emergency Plan.

Safety Analysis Report Page-8.0-0 Rev 50

11.5 Audits and Assessments Audits and Assessments LES will have a tiered approach to verifying compliance to procedures and performance to regulatory requirements.

ASSESSMENTS Assessments are owned and managed by the line organizations focused on effectiveness of activities and ensuring that IROFS, and any items that are essential to the function of IROFS, are reliable and are available to perform their intended safety functions. This approach includes performing Assessments on critical work activities associated with facility safety, environmental protection and other areas as identified via trends.

Assessments are performed to assure that facility activities are conducted in accordance with the written procedures and that the processes reviewed are effective. As a minimum, these assessments shall assess activities related to radiation protection, criticality safety control, hazardous chemical safety, industrial safety, fire protection, and environmental protection.

Personnel performing assessments do not require certification, but they are required to complete QA orientation training, as well as training on the assessment process. The nuclear criticality safety assessments are performed under the direction of the criticality safety staff.

Personnel performing these assessments do not report to the production organization and have no direct responsibility for the function or area being assessed. Assessments are conducted using approved procedures that meet the QAPD requirements. A schedule is established and maintained that identifies assessments to be performed and the responsible organization assigned to conduct the activity.

Deficiencies identified during the assessments requiring corrective action shall be forwarded to the responsible manager of the applicable area or function for action in accordance with the CAP procedure.

The Operations Group is assessed periodically to ensure that nuclear critical safety procedures are being followed and the process conditions have not been altered to adversely affect nuclear criticality safety. The frequency of these assessments is based on the controls identified in the NCS analyses and NCS evaluations. Assessments are conducted at least annually. In addition, weekly nuclear criticality safety walkthroughs of UF6 process areas are conducted and documented.

Assessment results are tracked and the data is periodically analyzed for potential trends.

Needed program improvements are identified to prevent recurrence and/or for continuous program improvements. The resulting trend is evaluated and reported to applicable management. This report documents the effectiveness of management measures in controlling activities, as well as deficiencies. Deficiencies identified in the trend report require corrective action in accordance with the applicable CAP procedure.

LBDCR-22-016 Assessments of nuclear criticality safety, performed in accordance with ANSI/ANS-8.19, will ensure that operations conform to criticality requirements. The nuclear criticality safety assessments are performed under the direction of the criticality safety staff. Personnel performing these assessments do not report to the production organization and have no direct responsibility for the function or area being assessed.

Safety Analysis Report Page-11.5-1 Rev 50

12.0 PHASED OPERATION General Accident Sequences EE-SEISMIC-WORKER EVAC IROFS39a FF-WORKER EVAC IROFS36a, 36d, &

36i, IROFS39b LBDCR-22-016 EE-CHEM RELEASE-WORKER EVAC IROFS39c EE-TORNADO MISSILE-SBM-CRDB SHELL & BUNKER WORKER IROFS39d Safety Significance Section 12.0 of the Safety Analysis Report has been initially established as an administrative change to describe the Phased Operation concept. There is no safety significance because none of the identified changes will be finalized and implemented until reviewed and approved in accordance with the LES configuration management program as described in § 11.1, Management Measures. Pursuant to 10 CFR 70.72, LES has established a system to evaluate, implement, and track each change to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel. Configuration management of IROFS, and any items that may affect the function of IROFS, is applied to all items identified within the scope of the IROFS boundary. All changes to structures, systems, equipment, components, and activities of personnel within the identified IROFS boundary are evaluated before the change is implemented. If the change requires an amendment to the License, Nuclear Regulatory Commission approval is received prior to implementation.

All proposed changes described in Section 12.0 are tracked and evaluated per the LES configuration management program prior to implementation. As the changes are processed, Section 12.0 will be revised to incorporate changes to the facility, processes, and programs.

Section 12.0 documents all site changes facilitated as a result of the Phased Operation approach.

Safety Analysis Report Page-12.0-2 Rev 50