ML22131A291

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Shine OL SER Chapter 6b.3 with No Open Items
ML22131A291
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Site: SHINE Medical Technologies
Issue date: 05/11/2022
From: Gavello M
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6b.4.3 Nuclear Criticality Safety The NRC staff evaluated the sufficiency of SHINEs nuclear criticality safety design criteria and methods, as described in SHINE FSAR Section 6b.3, Nuclear Criticality Safety, using the guidance and acceptance criteria from Section 6b.3 of the ISG augmenting NUREG-1537, Part

2. Where appropriate, the staff referenced the guidance contained in NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, (NRC, 2015) and NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, (NRC, 2001) to support the staffs evaluation of the ISG to NUREG-1537 acceptance criteria.

6b.4.3.1 Nuclear Criticality Safety Program General requirements to protect health and minimize danger to life and property in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.34(b) are implemented in practice by establishing and maintaining a nuclear criticality safety program (CSP) that assures subcriticality under normal and all credible abnormal conditions, with an approved margin of subcriticality for safety. This is typically implemented in conjunction with the double contingency principle (DCP) as stated in American National Standards Institute/American Nuclear Society (ANSI/ANS) 8.1-2014, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.

SHINE describes the CSP in SHINE FSAR Section 6b.3.1, Nuclear Criticality Safety Program, which states that SHINEs CSP applies to all nuclear processes within the Radioisotope Production Facility (RPF) and the Irradiation Facility (IF), excluding the Target Solution Vessels (TSVs).

Organization and Administration SHINE FSAR Section 6b.3.1 discusses the programs objectives and states that its goal is to ensure that workers, the public, and the environment are protected from the consequences of a criticality event. The CSP is executed by qualified staff using written procedures, which are maintained by SHINEs document control program.

SHINE FSAR Section 6b.3.1.1 discusses the organization of the CSP. SHINE has committed to meet, and has described a program whose structure is consistent with, the requirements of ANSI/ANS-8.1-2014 and ANSI/ANS-8.19-2014, Administrative Practices for Nuclear Criticality Safety, both of which are endorsed by NRC Regulatory Guide (RG)-3.71, Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores. SHINE FSAR Section 6b.3.1.1 discusses the responsibilities and roles of key program personnel, including the Chief Executive Officer, Safety Analysis Manager, SHINE facility management, fissionable material operation (FMO) supervisors, and nuclear criticality safety (NCS) staff. The overall responsibility of the CSP lies with the SHINE Chief Executive Officer; and the Safety Analysis Manager is the facility manager responsible for the programs implementation. NCS staff are kept administratively independent from operations to the extent practicable. The NRC staff finds that the information provided with respect to the CSP organization is consistent with the requirements of ANSI/ANS-8.1-2014 and ANSI/ANS-8.19-2014 and the ISG to NUREG-1537 acceptance criteria and are, therefore, acceptable.

SHINE FSAR Section 6b.3.1.2 discusses NCS staff training and qualifications. The NCS staff consists of three qualification levels: 1) NCS Analyst, 2) NCS Engineer, and 3) Senior NCS Engineer. The NCS training program consists of two tiers, with Tier 1 being directed toward personnel who manage, work in, or work near areas where a potential for criticality exists and 1

Tier 2 being specific to NCS staff. Tier 1 content is derived from ANSI/ANS-8.20-1991, Nuclear Criticality Safety Training, and Tier 2 content is derived from ANSI/ANS-8.26-2007, Criticality Safety Engineer Training and Qualification Program. Both tiers include content on procedural compliance, stop-work authority, response to criticality accident alarm system (CAAS) alarms, including evacuation to designated areas using designated routes, and reporting of defective or anomalous conditions. SHINE has committed to meet the requirements of ANSI/ANS-8.26-2007 for the training of NCS staff, and SHINEs specific training requirements for NCS staff are taken from this standard. The NRC staff finds that the information provided with respect to NCS staff training and qualifications are consistent with the ISG to NUREG-1537 acceptance criteria, as well as ANSI/ANS-8.20-1991, and ANSI/ANS-8.26-2007 (both endorsed by RG-3.71) and are, therefore, acceptable.

Management Measures Applied The management measures applied to the CSP consist of training, procedures (including NCS postings), and audits and assessments. These aspects of the CSP are discussed in SHINE FSAR Section 6b.3.1.

SHINE FSAR Section 6b.3.1.6 discusses NCS training. As previously discussed, SHINE states that it will establish and maintain a two-tiered NCS training program, consisting of content derived from ANSI/ANS-8.20-1991 directed toward those who manage, work in, or work near areas where the potential for criticality exists (Tier 1), and content derived from ANSI/ANS-8.26-2007 directed toward NCS staff (Tier 2). Both tiers include content on procedural compliance, stop-work authority, response to CAAS alarms including evacuation to designated areas using designated routes, and the reporting of defective or anomalous conditions. SHINE has committed to meet the requirements of ANSI/ANS-8.20-1991 and ANSI/ANS-8.26-2007, both endorsed by RG-3.71. The NRC staff finds that the information provided with respect to the NCS training are consistent with the requirements of ANSI/ANS-8.20-1991 and ANSI/ANS-8.26-2007 and the ISG to NUREG-1537 acceptance criteria and are, therefore, acceptable.

SHINE FSAR Section 6b.3.1.8 discusses NCS-related nonconformances. Personnel are trained to promptly report any deviations from procedures and/or unintended changes in process conditions to management. Nonconformances are entered into SHINEs corrective action program, investigated promptly, corrected as appropriate, and documented. Corrective actions are performed in accordance with procedures and with guidance from the NCS staff.

SHINE FSAR Section 6b.3.1.7 discusses NCS procedures (including NCS postings). As previously discussed, SHINE has committed to following the requirements of ANSI/ANS-8.19-2014 and ANSI/ANS-8.20-1991, which both discuss the use of procedures. SHINE FSAR Section 6b.3.1.7 states that activities involving fissile material are conducted using written and approved procedures. Section 6b.3.1.7 also states that for situations in which approved procedures are inadequate or do not exist, personnel are required to take no action until the NCS staff has evaluated the situation and provided instructions. To supplement procedures, SHINE also uses postings and labeling, as described in SHINE FSAR Section 6b.3.1.7. The NRC staff finds that the information provided is consistent with the ISG to NUREG-1537 acceptance criteria and, therefore, acceptable.

SHINE FSAR Section 6b.3.1.7 discusses CSP audits and assessments. SHINE performs several different types of NCS audits and assessments, which are consistent with the requirements of ANSI/ANS-8.19-2014 and have been incorporated into SHINEs Technical Specifications (TS). The audits and assessments consist of 1) annual audits of operations to 2

evaluate procedural compliance and process conditions, including walkthroughs of facility processes and procedures by NCS staff; 2) periodic reviews of active procedures by supervisors; 3) periodic reviews of procedural noncompliance and other NCS-related deficiencies; 4) annual reviews of nuclear criticality safety evaluations (NCSEs) to ensure their continued validity, with each NCSE and its associated calculations being reviewed at least once every three years; and 5) a triennial audit of the overall effectiveness of the CSP, with participation from SHINE management. The adequacy of NCS controls is routinely assessed by SHINEs audits and assessments. Deficiencies that are identified during an audit or assessment are promptly reported to management via the SHINE corrective action program, investigated promptly, corrected as appropriate, and documented. Action to correct deviations or alterations is taken in accordance with procedural requirements and with guidance obtained from the NCS staff. Action is taken to prevent recurrence for significant conditions adverse to quality. Records of NCS deficiencies and associated corrective actions are maintained in the corrective action program. The ISG augmenting NUREG-1537 contains an acceptance criterion that all aspects of the CSP will be audited at least every 2 years; however, SHINE states that it will conduct such audits triennially. In its response to the NRC staffs request for additional information dated January 29, 2021 (ADAMS Accession No. ML21029A102), SHINE stated that this longer audit frequency is justified based on its commitment to ANSI/ANS-8.19-2014, which states in paragraph 4.7 that, Management shall participate in auditing the overall effectiveness of the nuclear criticality safety program at least once every 3 years. The staff notes that RG-3.71 endorses the use of ANSI/ANS-8.19-2014. Therefore, based on SHINEs commitment to ANSI/ANS-8.19-2014, the staff determined that a triennial program audit is acceptable.

In addition to the management measures applied to the CSP, SHINE FSAR Section 6b.3.1.5 states that any process or design change that could impact NCS limits or controls is evaluated against the requirements of 10 CFR 50.59. Prior to implementing the change, the applicable NCSE is reviewed and revised, if necessary, to maintain the assurance of subcriticality under normal and all credible abnormal conditions with an approved margin of subcriticality for safety.

The NRC staffs evaluation of SHINEs change control process is discussed in Chapter 12 of this SER.

Use of Industry Standards SHINE FSAR Section 6b.3.1.3, Use of National Consensus Standards, states that SHINE commits to the requirements of the following ANSI/ANS standards, subject to the clarifications and exceptions provided in RG-3.71 (Rev. 3), with certain SHINE-specific limitations. The NRC staff notes that although the ISG augmenting NUREG-1537 acceptance criteria references different versions of these standards in certain instances, SHINEs commitments to these standards as discussed below are to the latest version of each standard consistent with Section 3.1 of NUREG-1537, Part 2, which states that design criteria should include references to applicable up-to-date, standards, guides, and codes. The staff also notes that in the context of ANSI/ANS standards, the use of the term requirements does not represent the regulatory requirements of 10 CFR Part 50 or 10 CFR Part 70; rather, it refers to the shall statements within the standard.

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

SHINE has committed to following the requirements of ANSI/ANS-8.1-2014. However, the clarification applied to this standard by RG-3.71 is related to subcritical limits for plutonium isotopes and is, therefore, not applicable to the SHINE facility.

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SHINE has committed to following the requirements of ANSI/ANS-8.3-1997 and acknowledges that the clarifications and exceptions applied to this standard by RG-3.71 are applicable to the SHINE facility.

  • ANSI/ANS-8.5-1996, Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material.

Borosilicate-glass Raschig rings are not used in the SHINE facility; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.6-1983, Safety in Conducting Subcritical Neutron-Multiplication Measurements In Situ.

SHINE has committed to following the requirements of ANSI/ANS-8.6-1983.

SHINE has committed to following the requirements of ANSI/ANS-8.7-1998.

  • ANSI/ANS-8.9-1987, Nuclear Criticality Safety for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials.

This standard was withdrawn by ANSI/ANS in May 1997 and is no longer maintained as an ANSI/ANS American National Standard; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.10-2015, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement.

SHINE does not rely on the criteria provided in this standard for determining the adequacy of shielding and confinement; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.12-1987, Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors.

Plutonium is not used as a fuel component at the SHINE facility; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.14-2004, Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors.

SHINE does not use soluble neutron absorbers for criticality control; therefore, this standard is not applicable to the SHINE facility.

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SHINE does not conduct operations with non-negligible quantities of selected actinides; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.17-2004, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors.

SHINE does not handle, store, or transport LWR fuel rods or units; therefore, this standard is not applicable to the SHINE facility.

SHINE has committed to following the requirements of ANSI/ANS-8.19-2014.

SHINE has committed to following the requirements of ANSI/ANS-8.20-1991.

  • ANSI/ANS-8.21-1995, Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors.

SHINE does not rely on the use of fixed absorbers as a means of criticality control; therefore, this standard is not applicable to the SHINE facility.

  • ANSI/ANS-8.22-1997, Nuclear Criticality Safety Based on Limiting and Controlling Moderators.

SHINE has committed to following the requirements of ANSI/ANS-8.22-1997.

SHINE has committed to following the requirements of ANSI/ANS-8.23-2007 and acknowledges that the clarification applied to this standard by RG-3.71 is applicable to the SHINE facility.

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

SHINE has committed to following the requirements of ANSI/ANS-8.24-2017 and acknowledges that the clarifications applied to this standard by RG-3.71 are applicable to the SHINE facility.

SHINE has committed to following the requirements of ANSI/ANS-8.26-2007.

SHINE does not possess irradiated LWR fuel assemblies; therefore, this standard is not applicable to the SHINE facility.

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The NRC staff determined that SHINE has committed to following the requirements of all relevant ANSI/ANS NCS standards. Therefore, the staff determined that the above commitments are consistent with the ISG augmenting NUREG-1537 acceptance criteria and are, therefore, acceptable.

Subcriticality and Double Contingency Principle SHINE FSAR Section 6b.3.1.3, Use of National Consensus Standards, states that SHINE commits to follow ANSI/ANS-8.1-2014, as endorsed by RG-3.71, which includes adherence to the double contingency principle (DCP) where practicable. Further, SHINE FSAR Section 6b.3.1.5, Computational System Validation, states that processes within the RPF generally comply with the DCP. This is expanded by SHINE FSAR Section 6b.3.2, Criticality Safety Controls, which states that control over two independent parameters is generally preferred over multiple controls over a single parameter, and if multiple controls on a single parameter are used, then a preference is given to diverse means of control. SHINE FSAR Section 6b.3.1.5 also states that the failure of a single control that maintains two or more controlled parameters is considered a single process upset with respect to the DCP. By letter dated December 10, 2020 (ADAMS Accession No. ML20357A087), SHINE stated that process upsets are considered credible unless they are not physically possible or are caused by a sequence of events involving many unlikely human actions or errors for which there is no reason or motive. SHINE further stated that the credibility of process upsets is based on the judgement of key professionals involved in the evaluation process from operations, design engineering, and safety analysis disciplines, considering factors such as the conditions of the system, its construction, and the applicable accident sequences. The NRC staff notes that the guidance provided in Appendix A to Chapter 5.0 of NUREG-1520 acknowledges that a qualitative approach to the DCP is acceptable. Therefore, the staff finds that SHINEs qualitative approach to evaluating the credibility and likelihood of process upsets is consistent with the DCP and guidance provided by Appendix A to Chapter 5.0 of NUREG-1520 and is, therefore, acceptable to the staff.

The NRC staff determined that the information is consistent with ANSI/ANS-8.1-2014 (endorsed by RG-3.71) and the ISG augmenting NUREG-1537 acceptance criteria and are, therefore, acceptable to the staff.

Technical Practices SHINE describes the NCS technical practices in SHINE FSAR Sections 6b.3.1.4, Nuclear Criticality Safety Evaluations, 6b.3.1.5, Computational System Validation, and 6b.3.2, Criticality Safety Controls. This includes the performance and documentation of NCSEs, the treatment of NCS parameters and their methods of control, the derivation of NCS limits, computational system validation, and the establishment and use of a margin of subcriticality for safety.

SHINE FSAR Section 6b.3.1.4 discusses SHINEs performance and documentation of NCSEs.

NCSEs are conducted for each FMO to ensure subcriticality under normal and all credible abnormal conditions with an approved margin of subcriticality for safety, with the exception of the Irradiation Unit (IU) cells and material staging building (see Section 6b.4.3.2 of this SER for further discussion of criticality hazards in the IU cells and material staging building and SHINEs request for exemption from the requirements of 10 CFR 70.24 for these areas). For analytical purposes, all fissionable isotopes are conservatively assumed to be fissile. NCSEs are conducted using hazard evaluation techniques, including What-if, What-if Checklist, and Event Tree Analysis, to identify potential accident sequences leading to inadvertent criticality.

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When the double contingency principle is used, the NCSEs also describe how the double contingency principle is implemented. SHINE will perform NCSEs using the industry-accepted and peer-reviewed methods discussed in the requirements of ANSI/ANS-8.1-2014 and ANSI/ANS-8.19-2014, both endorsed by RG-3.71.

NCS limits identified in NCSEs are derived using one of three methods: 1) industry-accepted and peer-reviewed references including ANSI/ANS standards, 2) hand calculations using industry-accepted and peer-reviewed techniques consistent with their limitations, or 3) computational methods. Limits are derived by assuming optimum or most-reactive credible values unless otherwise controlled. For cases in which less than optimum values are used, the basis is documented in the appropriate NCSE. Operating limits are derived conservatively in consideration to any potential process variability and uncertainty to ensure NCS limits are unlikely to be exceeded. When computational methods (i.e., code) are used to derive NCS limits and/or perform NCS analyses, it is used consistent with the limitations and penalties identified in the codes validation report. SHINE TS Table 5.5.4, Controls, states that engineered criticality controls are identified in the criticality safety evaluations to prevent criticality in the SHINE Facility, excluding the TSVs.

SHINE FSAR Section 6b.3.2 discusses the various NCS control methods, including preference for engineered controls over administrative controls, and passive engineered controls over active controls. Controls are established to restrict certain NCS parameters within subcritical limits. The NCS parameters SHINE controls are mass, moderation, enrichment, geometry, volume, concentration, interaction, physiochemical form, reflection, heterogeneity, density, and process variables. SHINE does not use fixed or soluble neutron absorbers as a method of control, and therefore has not provided any information on the treatment of absorption as an NCS parameter. The discussion associated with the treatment of each of these parameters is below.

The NRC staff notes that several of the ISG augmenting NUREG-1537 acceptance criteria under one or more of the individual parameters are covered by the more general discussions in SHINE FSAR Section 6b.3.1. For example, the expectation that instrumentation is relied on to verify compliance with limits on mass, density, enrichment, etc., will be subjected to facility management measures is bound by a general statement in SHINE FSAR Section 6b.3.2. The expectations that firefighting procedures will be evaluated for moderator intrusion, or that all precipitating agents will be identified and controlled against, are corollary to the general requirement that the licensee ensure that processes are subcritical under normal and all credible abnormal conditions, as well as SHINEs commitment to follow the requirements of ANSI/ANS-8.22-1997. Regarding the ISG augmenting NUREG-1537 acceptance criteria stating that process variables that can affect the value of a particular parameter should be controlled by safety related controls identified in NCSEs, the staff finds it sufficient for the licensee to follow its SHINE Safety Analysis (SSA) methodology as described in SHINE FSAR Section 13a2 in determining what controls should be designated as safety related controls. The staffs evaluation of SHINEs SSA methodology is discussed in Chapter 13 of this SER.

The NRC staff noted that several subsections of SHINE FSAR Section 6b.3.2 pertaining to NCS parameters contain provisions which are simply definitions of the parameter or state that the parameter may be used on its own in combination with other parameters. In accordance with the guidance provided in Appendix A to Chapter 5.0 of NUREG-1520, this is acceptable in the context the DCP and is, therefore, acceptable to the NRC staff. The above discussion is applicable to each of the parameters discussed in the subsections of SHINE FSAR Section 7

6b.3.2. Significant points regarding the specific parameters are discussed in the paragraphs below.

Mass: SHINE states that whenever mass limits are based on assuming a certain weight percent of uranium, either the entire mass present will be ascribed to uranium or the actual weight percent determined by physical measurement. Thus, any material associated with a special nuclear material (SNM) process will be treated as having a high uranium content until demonstrated otherwise. SHINE also assumes a conservative process density to calculate mass when the dimensions of equipment or containers with fixed geometry are used to limit mass, and to demonstrate that the largest mass resulting from a single failure remains subcritical wherever over-batching is credible. The use of instrumentation to measure mass is addressed by a general statement that when measurement of a parameter is needed, instrumentation subject to facility management measures is used. The NRC staff finds that the information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Geometry and Volume: SHINE restricts SNM volume with geometry and to verify all dimensions relied on in demonstrating subcriticality before beginning operations, in response to changes in operations, and at periodic intervals. Relevant dimensions and material properties are maintained by the facilitys configuration management program. Abnormal conditions involving the loss of geometry are evaluated in NCSEs, if credible. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Density: SHINEs discussions regarding the treatment of density as a controlled parameter are provided by general statements that apply to all other controlled parameters. Limits on density are established with consideration given to any tolerances and uncertainty. If control is based on measuring density, independent means of measurement are used subject to facility management measures. If process variables can affect the normal or most reactive credible value density, controls to maintain density within a certain range are used. For cases where a single parameter limit on density is used, all other parameters are evaluated at their optimum or most reactive credible value. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Enrichment: SHINE assumes a bounding enrichment for all fissile material based on a facility-wide maximum authorized enrichment controlled by receipt inspections of feed material.

Measurements of the enrichment of feed material are performed using instrumentation subject to facility management measures. SHINE does not have any processes capable of enriching SNM further. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Reflection: SHINE considers wall thickness and all adjacent reflecting material when determining subcritical limits for individually evaluated units of fissile material. Although SHINE does not explicitly state that individually evaluated units will be farther than 12 inches away from reflecting adjacent materials, SHINE will establish and document criteria for determining whether materials are sufficiently spaced to be considered neutronically isolated. When reflection is not controlled, full reflection is assumed and is represented by 12 inches of tight-fitting water or 24 inches of tight-fitting concrete, as appropriate. Unit arrays are evaluated using the most reactive combination of interstitial moderation and exterior array reflection.

Minimum reflection conditions, equivalent to a 1-inch tight-fitting water reflector, are assumed to account for personnel and other transient reflectors not explicitly included in criticality 8

calculations. When less than full reflection is assumed, controls to limit reflection around individual units are established, with rigid barriers preferred. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Moderation: SHINE commits to following the requirements of ANSI/ANS-8.22-1997. Physical structures are designed to prevent the ingress of moderators, and conspicuously marked moderation-controlled areas are used to exclude moderators from process areas. Firefighting procedures for use in moderation-controlled areas, as well as the effects of fire and the activation of fire suppression systems, are evaluated in NCSEs, and restrictions are applied to the use of moderating firefighting agents. The use of instrumentation to measure moderation is addressed by a general statement that when measurement of a parameter is needed, instrumentation subject to facility management measures is used. Process variables that can affect moderation are also addressed by a general statement to identify in accident analyses the process variables relied on to control or monitor controlled parameters, with sufficient management measures applied. The associated parameter (in this case moderation) is explicitly identified and the correlation of process variables to the associated parameter is established by experiment or plant-specific measurements. When moderation needs to be sampled, dual independent sampling methods are used. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Concentration: SHINE implements controls to limit concentration unless the process has been demonstrated to be subcritical under the most reactive credible uranium concentrations. For solution tanks under concentration control, precautions are taken to preclude the introduction of precipitating agents, including keeping the tank closed and locked. When concentration needs to be sampled, such as prior to a transfer of solution to an unfavorable geometry tank, dual independent sampling methods and/or in-line monitoring such that no single error may result in the transfer of concentrated solution are used. Process variables that can affect the solubility of fissile solutions are controlled and monitored, and the need to ensure homogeneity of solution is assessed in NCSEs. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Interaction: SHINE maintains physical separation between fissile-bearing units with engineered controls. If engineered controls are not feasible, visual aids are used to support administrative controls. The structural integrity of spacers, storage racks, etc. is sufficient to ensure subcriticality under normal and all credible abnormal conditions, including seismic events.

Movable engineered devices are periodically inspected to verify that no deformation has occurred. The NRC staff finds that this information is consistent with the ISG augmenting NUREG-1537 acceptance criteria and, therefore, acceptable.

Physiochemical Form and Process Variables: Physiochemical form and process variables are not explicitly discussed in the ISG augmenting NUREG-1537 acceptance criteria, but rather are used to indirectly control one or more NCS parameter. For example, specifying the material form as uranyl nitrate solution may be necessary to use certain dimensional or mass limits, and implicitly takes credit for the neutron absorbing properties of nitrogen. As another example, specifying the form as uranium dioxide powder implicitly assumes a maximum enrichment and density. Reliance on this parameter is based on known scientific principles or known physical or chemical properties, in conjunction with experimental data supported by operating history. By letter dated December 10, 2020, SHINE stated that correlation of process variables to controlled parameters is accomplished by direct or indirect measurements. Direct correlations apply when 9

monitored process variables are directly measured. Indirect correlations may be established when direct measurements of controls are not possible and may be developed using known empirical or theoretical relationships or developed through plant-specific data collection.

Correlation inputs will be defined and documented for a specific controlled parameter, including process assumptions and characteristics. The NRC staff finds that this is sufficient to ensure that reliance is not placed on as-found conditions or on process assumptions and characteristics that are not appropriately controlled. SHINE will establish controls to limit material composition to a particular form. Process variables that can change fissile material composition to a more reactive physiochemical form are identified as controls, and both in-situ changes in physiochemical form and the migration of material between process areas are considered in evaluating credible abnormal conditions. While the ISG augmenting NUREG-1537 does not contain acceptance criteria specifically associated with these parameters, the staff finds that this information is consistent with the general principles that apply to all parameters as discussed above and, therefore, acceptable.

Heterogeneity: The ISG augmenting NUREG-1537 does not provide specific acceptance criteria regarding the treatment of heterogeneity. Rather, it simply states that heterogeneous effects should be considered as appropriate. Given the enrichment and material forms of SHINEs processes, heterogeneity effects are necessary to consider. Therefore, the acceptance criteria from NUREG-1520 was used as guidance in determining whether the information provided regarding heterogeneity are acceptable. SHINE evaluates potential methods of causing fissile solution to become inhomogeneous and establish controls as necessary. If heterogeneity is considered credible, its effect is evaluated in NCSEs.

Assumptions that can affect the physical scale of homogeneity are based on observed physical characteristics, and process variables that can affect the scale of heterogeneity are controlled.

Given that SHINEs methods for determining subcritical limits are based on comparison with uranyl fluoride and uranyl nitrate systems, and that SHINEs computational method for performing NCS analyses was validated, in part, against uranyl fluoride and uranyl nitrate benchmark experiments, the NRC staff performed a literary search and an independent analysis to assess the potential impacts of heterogeneity as both uranyl fluoride and uranyl nitrate present special concerns due to the formation of water hydrates at low hydrogen-to-uranium ratios (H/U). Oak Ridge National Laboratory (ORNL) report ORNL/TM-12292, Estimated Critical Conditions for [Uranyl Fluoride - Water] Systems in Fully Water-Reflected Spherical Geometry, suggests that special considerations should be given to uranyl fluoride systems with a H/U ratio less than 4.0 and to uranyl nitrate systems with an H/U less than 12. A review of ARH-600, Criticality Handbook, Volume I, suggests that uranyl sulfate may be subject to the same phenomena and require special considerations for H/U less than 7. Below these values, the formation of water hydrates complicates efforts to determine the bulk fissile material density as concentration values can be somewhat erroneous. Above these values, these effects are not of concern, and a general molar volume additive approach can be used. As stated in SHINEs validation report, CALC-2018-0012, the H/U ratio for uranium solutions in SHINE processes that are evaluated using SHINEs computational method are all well above these limits. Although SHINE does have processes in which low H/U ratios may occur (e.g., the dissolution of uranium oxides in sulfuric acid to form uranyl sulfate solution), SHINE assumes the most reactive material composition in analyses. In this particular example, the process of dissolving uranium oxide into uranyl sulfate solution would be evaluated assuming the most reactive composition (moderated uranium oxide) for the entire process. This negates the need to consider the potential effects of low H/U ratios involving uranyl sulfate. Additionally, the application of subcritical limits (SPLs) derived from ANSI/ANS standards for uranium metal and oxides negates the need to conduct an analysis to identify the optimum solution concentration 10

as the SPLs already inherently assume optimum. In cases where optimum solution concentration does need to be determined, such as in the design of a favorable geometry vessel or in accident sequences involving precipitation, the optimum concentration for a 20 weight percent U-235 uranyl sulfate system corresponds to an H/U significantly greater than the H/U values listed above for the formation of water hydrates. This alleviates any concern of skewed results due to the formation of water hydrates. Based on the NRC staffs literary search and independent analysis, the staff determined that SHINEs practices for controlling heterogeneity appropriately bound any potential concerns regarding the formation of water hydrates at low H/U ratios. While the ISG augmenting NUREG-1537 does not contain acceptance criteria specifically associated with this parameter, the staff finds that this information is consistent with the general principles that apply to all parameters as discussed above and, therefore, acceptable.

SHINE discusses the validation and verification of computational methods in SHINE FSAR Section 6b.3.1.5, Computational System Validation. SHINE has committed to following the requirements of ANSI/ANS-8.24-2017 and has stated that computational systems are verified and validated using the guidance contained in NUREG/CR-6698. SHINE validated its computational methods by comparing the results calculated with computer models to the results of 128 benchmark experiments from the Handbook of the International Criticality Safety Benchmark Evaluation Project (ICBEP). Benchmarks were selected for evaluation based on their similarity to SHINE solution systems, and a modified form of the Shapiro-Wilk test for normality from NUREG/CR-6698 was used to determine whether the resulting benchmark data was normally distributed. The single-sided tolerance limit approach from Section 2.4.4 of NUREG/CR-6698 was used to calculate the upper subcritical limit (USL), which is defined in SHINE FSAR Section 6b.3.1.5 as the difference between unity and the sum of the bias, the bias uncertainty, and minimum margin of subcriticality (MMS). For cases involving positive bias, SHINE conservatively assumes the bias to be zero. SHINE FSAR Section 6b.3.1.5 states that verification of the Monte Carlo N-Particle software installation was performed using developer-supplied verification tools, and that re-verification of the computational system is conducted following any changes to the hardware or operating system. The NRC staff finds that this information is consistent with the ISG to NUREG-1537 acceptance criteria and are, therefore, acceptable.

SHINEs discussion regarding the use of a margin of subcriticality for safety (MMS) is in SHINE FSAR Section 6b.3.1.5. SHINEs computational method identified in SHINE document CALC-2018-0012, MCNP5 Validation for Reactivity in Solution Systems for the SHINE Facility, is MCNP5 using the ENDF/B-VII.1 standard cross-section library and the ENDF/B.VII.0 thermal scattering cross-section library, which is validated to establish the areas of applicability in which the code can be used.

SHINE relies on the use of SPLs and other data from NRC-endorsed ANSI/ANS standards to establish safe limits for non-solution processes. For solution processes, the TSV dump tank and the TSV off-gas system (TOGS), a margin of 0.05 is applied with an additional penalty of 0.01 to account for any differences between benchmark material composition and the material composition of SHINE processes. This results in a margin of subcriticality of 0.06, which corresponds to an upper subcritical limit (USL) of 0.94. Appendix B to Chapter 5.0 of NUREG-1520 states that a margin of subcriticality of 0.05 is generally acceptable for low-enriched uranium facilities provided 1) a criticality code validation study has been performed consistent with ANSI/ANS-8.24-2017 and/or NUREG/CR-6698, 2) there is an acceptable number of critical experiments with similar geometric forms, material compositions, and neutron energy spectra to the applicable processes, and 3) the processes evaluated include materials and process 11

conditions similar to those that occur in low-enriched fuel cycle applications. As previously discussed, SHINE has performed a validation study consistent with NUREG/CR-6698 and ANSI/ANS-8.24-2017 involving 128 benchmark experiments from the ICBEP selected for evaluation based on their similarity to SHINE solution systems. Although SHINEs processes involve some operations that are not typical of a fuel cycle facility (e.g., irradiation of fissile solution), such processes are limited to the TSVs (TSVs are not subject to the SHINE CSP).

The majority of SHINEs processes are largely similar to that of a fuel cycle facility.

Furthermore, SHINE has applied a penalty to the USL to account for any differences between benchmark material composition and the material composition of SHINEs processes. Appendix B to Chapter 5.0 of NUREG-1520 states that the justification of a margin of subcriticality should be based on 1) conservative practices in calculational models, 2) validation methodology and results, and 3) additional risk informed considerations. Given that SHINE uses conservative NCS practices (see Section 6b.4.3.1 of this SER), and has performed a validation study consistent with NUREG/CR-6698 and ANSI/ANS-8.24-2017 using a sufficient quantity of data from a quality source (ICBEP) with an acceptable statistical methodology and benchmarks similar to SHINE processes; the NRC staff determined that a margin of subcriticality of 0.06 (corresponding to a USL of 0.94) is acceptable.

6b.4.3.2 Criticality Accident Alarm System In accordance with 10 CFR 50.68(a), SHINE is required to comply with 10 CFR 70.24 or meet certain criteria described in 10 CFR 50.68(b). SHINEs discussions regarding the use of a criticality accident alarm system (CAAS) are in SHINE FSAR Section 6b.3.3, Criticality Accident Alarm System. SHINE FSAR Section 6b.3.3 states that the SHINE facility maintains CAAS coverage in areas where quantities of fissile material greater than the limits identified in 10 CFR 70.24(a) are used, handled, or stored, except for those areas exempt from the requirements of 10 CFR 70.24 (IU cells and the material storage building). Section 6b.3.3 further states that the CAAS is designed to meet the requirements of 10 CFR 70.24 and conforms to the requirements of ANSI/ANS-8.3-1997, as endorsed by RG-3.71.

SHINE states in SHINE FSAR Section 6a2.3.2, Criticality Accident Alarm System, that coverage of SNM in the IU cells is provided by the neutron flux detection system (NFDS) and level instrumentation in the TSV dump tank, which provides indication of abnormal conditions in the IU cells.

Criticality Accident Alarm System SHINE FSAR Section 6b.3.3 states that the CAAS is designed to meet the requirements of 10 CFR 70.24. Section 6b.3.3 further states that the CAAS conforms to the requirements of ANSI/ANS-8.3-1997, as endorsed by RG-3.71. Despite this being a different version of RG-3.71 than the Revision 1 version referenced in the ISG augmenting NUREG-1537 acceptance criteria, the NRC staff finds this acceptable as both versions endorse ANSI/ANS-8.3-1997 with the following exceptions:

  • at or above the mass thresholds established in ANSI/ANS-8.3-1997, Section 4.2.1, the applicant should commit to CAAS coverage in each area where SNM is handled, stored, or used - as opposed to merely an evaluation for such areas;
  • two detectors are required for each area requiring CAAS coverage - as opposed to coverage by a single reliable detector; and
  • the CAAS is required to be capable of detecting a criticality condition that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an 12

unshielded distance of 2 meters from the reacting material within 1 minute - as opposed to requiring CAAS coverage in areas where personnel would be subject to excessive radiation dose, defined as any dose corresponding to a neutron and gamma combined absorbed dose equal to or greater than 12 rads in free air.

The NRC staff finds that this criteria, along with the requirement to meet the criticality accident requirements of 10 CFR 70.24 in SHINE FSAR Section 6b.3, satisfies the ISG augmenting NUREG-1537 acceptance criteria associated with the applicant maintaining a CAAS that provides two detector coverage and is capable of detecting a criticality condition that produces a combined absorbed dose in soft tissue of 20 rads at an unshielded distance of 2 meters within 1 minute.

To accommodate maintenance and testing activities, SHINE FSAR Section 6b.3.3.2, Criticality Accident Alarm System Design, states that the CAAS detectors are arranged such that all areas within the main production facility requiring coverage generally receive coverage from at least three detectors, allowing a single detector for any given area to be taken out of service for a specified period of time. For maintenance or testing evolutions requiring multiple detectors or the logic unit to be taken out of service, SHINE implements administrative controls to secure the movement of fissile material and limit personnel access (i.e., quarantine) to affected areas until CAAS coverage is restored. Portable instruments may be used in rare circumstances. SHINE FSAR Section 6b.3.3.2 states that the CAAS is designed to be resistant from credible events, such as fire, explosion, a corrosive atmosphere, seismic shock, and other adverse conditions that do not result in evacuation of the entire facility. SHINE FSAR Section 6b.3.3.2 further states that the CAAS will energize clearly audible alarm signals if accidental criticality were to occur. The NRC staff finds that this information is consistent with the acceptance criteria of the ISG augmenting NUREG-1537; therefore, the staff finds it acceptable.

SHINE FSAR Section 6b.3.3.1 states that the minimum accident of concern (MAC) was developed based on a critical sphere of 20 wt% U-235 uranyl sulfate solution. Transport analysis was used to convert the neutron and gamma spectrum of the MAC to a point source, which was then used to determine the appropriate detector placement based on the facility structure, shielding, and potential intervening equipment. The thresholds defined in 10 CFR 70.24 were used to establish detection thresholds. Neutron detectors were selected for use to reduce potential inappropriate interference from multiple gamma sources throughout the facility.

NRC staff determined that although processes at the SHINE facility involve uranium compositions other than uranyl sulfate solution (such as uranium metal and oxides), uranyl sulfate is the uranium composition for the majority of SHINEs processes. Additionally, the information contained in Los Alamos National Labs LA-13638, A Review of Criticality Accidents, demonstrates that 21 of the 22 historical process-related criticalities involved solutions and suggests that solutions (such as uranyl sulfate) represent the uranium compositions most likely to be involved in a process-related criticality accident. NRC staff determined that the use of neutron detectors is appropriate given the potential for interference from gamma sources throughout the SHINE facility. NRC staff determined that SHINEs CAAS is appropriate for the facility and type of radiation detected, the intervening shielding, and the magnitude of the accident of concern. The staff therefore finds this information to be acceptable.

Request for Exemption from 10 CFR 70.24(a) 13

In accordance with 10 CFR 50.68(a), SHINE is required to comply with 10 CFR 70.24 or meet certain criteria described in 10 CFR 50.68(b). SHINE FSAR Section 6b.3.3 states that SHINEs CAAS is designed to meet the requirements of 10 CFR 70.24. In letter dated January 29, 2021 (ADAMS Accession No ML21029A038), SHINE requested an exemption from the requirements of 10 CFR 70.24(a) for two areas: 1) the IU cells within the irradiation facility, and 2) the material staging building (MATB).

In accordance with 10 CFR 70.17(a), the Commission may grant exemptions, provided the exemptions: 1) are authorized by law, 2) will not endanger life or property or the common defense and security, and 3) are otherwise in the public interest. The NRC staff evaluated SHINEs request for exemption from the requirements of 10 CFR 70.24(a) for the IU cells and the MATB.

IU Cells: For the IU cells during normal operation, SHINE stated that the cells are a robust shielded enclosure that is designed to protect workers and the public from the irradiation operation. SHINE further stated that SNM handling within the IU cells occurs in a light water pool, which provides radiation shielding during irradiation operations, as well as would provide protection from radiation due to an accidental criticality. The components within the cell are designed to be subcritical based on control of geometry with no credible means of deformation or loss of credited geometric properties. Furthermore, SHINE stated that the NFDS would detect the minimum accident of concern if criticality were to occur in the TSV dump tank, presenting as an increased count rate visible to operators through the process integrated control system. Inadvertent criticality in the TOGS would require a precursor condition to occur (i.e.,

solution overflowing into the TOGS), which would be detected by safety-related high-level indicators alerting operations personnel of the need to take appropriate response actions.

The NRC staff determined that although a credible criticality hazard exists in the IU cells, appropriate measures are in place to limit the risk of criticality and protect workers, the public, and the environment should criticality occur. The staffs evaluation of shielding is discussed in Chapter 4 of this SER. The staff determined that an exemption from the requirements of 10 CFR 70.24(a) for the IU cells would not endanger life or property or the common defense and security. The staff determined that this exemption is in the public intertest because it would enable SHINE to safely produce medical isotopes without having to maintain a CAAS in the IU cells, unnecessarily exposing maintenance personnel to radiation doses. Additionally, the staff determined that the requested exemption is permissible under the Atomic Energy Act of 1954, as amended, and other regulatory requirements. Therefore, the staff determined that the requested exemption from the requirements of 10 CFR 70.24(a) for the IU cells is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

Material Staging Building: The MATB provides a location for packaged radioactive material, both as-generated solid waste and solidified liquid waste, to decay until it can be transported to an off-site final disposal location. In its exemption request, SHINE stated that the material in the MATB is as-generated solid wastes packaged and staged for transport that meets the requirements of 10 CFR 71.15(a), as well as solidified liquid waste stored in packages that meet the requirements of 10 CFR 71.15(c). The NRC staff reviewed NUREG/CR-7239, Review of Exemptions and General Licenses for Fissile Material in 10 CFR 71, Sections 4.1.1 and 4.1.3, which describe the basis for the criteria in 10 CFR 71.15(a) and (c) to consider transportation packages exempt from the criticality safety requirements of 10 CFR Part 71. The staff noted that these transportation exemptions do not apply to SNM being stored under 10 CFR Part 50 or 10 CFR Part 70, including 10 CFR 70.24(a). However, the staff reviewed the NUREG/CR-7239 14

and SHINEs submittal to determine if a technical basis exists to use these criteria as justification for exemption from the requirements of 10 CFR 70.24(a).

10 CFR 71.15(a): The transportation exemption criteria in 10 CFR 71.15(a) is for individual packages containing 2 grams or less of fissile material. Section 4.1.1 of NUREG/CR-7239 states that criticality would not be credible based on practical and economic considerations for such packages. Specifically, Section 4.1.1 of NUREG/CR-7239 states that a cubic array of 84,853 one-liter packages, each containing 2 grams of U-235 with no absorbers or packaging material at near-optimal moderation, is required for criticality to be possible. The NRC staff determined that this basis does not apply to the MATB because the purpose of the MATB is to store many such packages, and the MATB is not physically restricted to less than 84,853 liters.

However, the staff determined that similar logic can be applied in that packages that meet 10 CFR 71.15(a) requirements stored in the MATB would contain as-generated solid waste absent of interstitial moderation. Packages containing, or potentially containing, interstitial moderation would not satisfy the conditions for treatment as solid waste and, therefore, would likely be treated as solidified liquid waste and packaged to meet 10 CFR 71.15(c) or further processed to remove the presence of interstitial moderation. The argument discussed in NUREG/CR-7239, which establishes that a minimum of 84,853 one-liter packages would be required for criticality, assumes near-optimal moderation with no absorbers or packaging material. Absent a significant amount of moderating material, criticality would not be possible. In order to introduce a significant amount of moderating material into the MATB, a number of significant, difficult, and unauthorized changes in process conditions would need to occur. The NRC staff considers such a condition to be not credible short of a willful, concerted effort. In addition to the required presence of significant amounts of moderating material, the packages would need to be arranged in a specific geometrical configuration for criticality to occur. Each package is limited to 2 grams fissile material and spacing between the fissile material in each package is provided by packaging material. The staff considers a geometrical arrangement that could support criticality to be contrived and not credible short of several unauthorized, willful changes in process conditions. Given the required presence of significant amounts of moderating material and the specific geometrical configuration required, the staff determined that the conditions required for inadvertent criticality to occur in the MATB involving packages that meet 10 CFR 71.15(a) are highly contrived and are extremely unlikely short of an unauthorized, concerted effort. The staff determined that packages meeting 10 CFR 71.15(a) do not present a credible criticality hazard, and an exemption from the requirements of 10 CFR 70.24(a) for such packages stored in the MATB is, therefore, warranted.

10 CFR 71.15(c): The transportation exemption criteria in 10 CFR 71.15(c) is for packages of large volumes of low-concentration, solid fissile material commingled with solid non-fissile material. The quantity of fissile material is not limited, but it must be an essentially homogeneous mixture of fissile and non-fissile material such that no more than 180 grams of fissile material is distributed within 360,000 grams of solid non-fissile material. This 2000:1 ratio represents approximately 60% of the minimum critical fissile material concentration of 1.33 g 235 U/L in a 1600-g-SiO2/L matrix (NUREG/CR-7239 Section 4.1.3). The NRC staff determined that the technical basis for this exemption criteria applies to material being stored at the MATB as it also involves large volumes of diluted dry materials packaged in accordance with 10 CFR 71.15(c). The staff determined that packages meeting 10 CFR 71.15(c) do not present a credible criticality hazard, and an exemption from the requirements of 10 CFR 70.24(a) for such packages stored in the MATB is, therefore, warranted.

The NRC staff determined that no credible criticality hazard exists in the MATB. Therefore, the staff determined that an exemption from the requirements of 10 CFR 70.24(a) for the MATB 15

would not endanger life or property or the common defense and security. The staff determined that this exemption is in the public intertest because it would enable SHINE to safely produce medical isotopes without having to maintain a CAAS in the MATB, unnecessarily exposing maintenance personnel to radiation doses. Additionally, the staff determined that the requested exemption is permissible under the Atomic Energy Act of 1954, as amended, and other regulatory requirements. Therefore, the staff determined that the requested exemption from the requirements of 10 CFR 70.24(a) for the MATB is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

Emergency Planning and Response SHINE FSAR Section 6b.3.1.3, Use of National Consensus Standards, states that SHINE commits to the requirements of ANSI/ANS-8.23-2007, as endorsed by RG-3.71. SHINE FSAR Section 6b.3.1.8.1, Planned Response to Criticality Accidents, states that SHINE maintains an emergency plan which includes the planned response to criticality accidents. SHINE FSAR Section 6b.3.1.8.1 further states that the emergency plan contains information on the provision of personnel accident dosimeters in areas that require CAAS coverage, arrangements for on-site decontamination of personnel, and the transport and medical treatment of exposed individuals. The SHINE Emergency Plan, Section 8.6.2, Assembly, states that SHINE has the capability of quickly identifying individuals who have received doses of 10 rads or more due to a criticality accident via reading of electronic dosimeters worn by personnel in the radiological controlled area. The SHINE Emergency Plan, Section 9.6, Personnel Monitoring Equipment, states that SHINE maintains emergency dosimetry that is readily available to emergency support personnel, and that equipment for prompt onsite readouts is maintained onsite. Fixed criticality accident dosimeters or instruments are located within the facility to provide spectrum information and assist in the reconstruction of a criticality accident. The SHINE Emergency Plan, Section 11.7, Emergency Plan and Procedure Use and Maintenance, states that SHINE maintains implementing procedures for the emergency plan for each area in which SNM is handled, used, or stored to ensure that all personnel evacuate to designated areas in the event of a CAAS alarm. The SHINE Emergency Plan, Sections 3.3.6, Criticality Safety Engineer, 3.5.2, Criticality Safety Lead Engineer, and 11.1.6, Criticality Safety Engineer, state that qualified NCS engineers are responsible for advising and assisting the emergency organization in response to a criticality event, and that they are trained on their responsibilities. The NRC staff finds that this information is consistent with the acceptance criteria of the ISG augmenting NUREG-1537; therefore, the staff finds it acceptable.

SHINE FSAR Section 6b.3.3.2, Criticality Accident Alarm System Design, states that the CAAS is equipped with a backup connection to the uninterruptible electrical power supply system (UPSS) and batteries to keep the system in operation for at least two hours following a facility loss of off-site power, providing operators with sufficient time to secure the movement of fissile material before a loss of alarm system coverage occurs.

SHINE FSAR Section 6b.3.3.2, states that for maintenance or other conditions which would disable multiple detectors or the logic unit, administrative controls are used to secure the movement of fissile material and limit personnel access to the affected areas until alarm system coverage is restored. The FSAR further describes that these administrative controls are specific to the various processes within the radioisotope production facility (RPF) and include short time allowances to restore the system to full operation in lieu of immediate process shutdown in areas where process shutdown creates additional risk to personnel.

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The NRC staff determined that information provided for the CAAS, and emergency planning and response provide reasonable assurance of adequate protection against the consequences of a criticality accident in accordance with the requirements of 10 CFR 70.24(b).

Reporting Requirements SHINE FSAR Section 6b.3.1.8, Criticality Safety Nonconformances, states that NCS events are reported to the NRC in accordance with the reporting requirements of 10 CFR 70.50, 10 CFR 70.52, and Appendix A to 10 CFR Part 70. However, because SHINE does not have an integrated safety analysis and associated controls designated as items relied on for safety (IROFS) pursuant to 10 CFR 70.61, modifications to the reporting requirements of Appendix A to 10 CFR Part 70 are specified in the SHINE TS. TS Section 5.8.3, Additional Event Reporting Requirements, states that the following events shall be reported to the NRC Operations Center within one hour of discovery, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 60 days:

  • an inadvertent nuclear criticality;
  • an acute intake by an individual of 30 mg or greater of uranium in a soluble form;
  • an acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that could endanger the life of a worker or could lead to irreversible or other serious, long-lasting health effects to any individual located outside the owner controlled area; and
  • an event or condition such that no credited controls, as documented in the SSA, remain available and reliable, in an accident sequence evaluated in the SSA.

TS Section 5.8.3 further states that the following events shall be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 60 days:

  • any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that described in the SSA, and which results in inadequate controls in place to limit the risk of chemical, radiological, or criticality hazards to an acceptable risk level, as required by the SSA;
  • loss or degradation of credited controls, as documented in the SSA, other than those items controlled by a limiting condition of operation established in Section 3 of the SHINE Technical Specifications, that results in a failure to limit the risk of chemical, radiological, or criticality hazards to an acceptable risk level, as required by the SSA;
  • an acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed materials that could lead to irreversible or other serious, long-lasting health effects to a worker, or could cause mild transient health effects to any individual located outside the owner controlled area; and
  • any natural phenomenon or other external event, including fires internal and external to the facility, that has affected or may have affected the intended safety function or availability or reliability of one or more safety-related structures, systems, or components.

TS Section 5.8.2, Special Reports, describes the special reports that will be submitted to the NRC. Special reports will be reported no later than the following working day by telephone and confirmed in writing by facsimile or similar conveyance to the NRC Operations Center, to be followed by a written report to the NRC Document Control Desk describing the circumstances of 17

the event within 14 days. This includes reports for observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to operations, as well as reports for safety system component malfunctions that render or could render the safety system incapable of performing its intended safety function.

The NRC staff determined that the above information is consistent with, and meets the intent of, the reporting requirements of Appendix A to 10 CFR Part 70. The staff determined that the above information is sufficient to provide reasonable assurance that events constituting a required report to the NRC will be reported appropriately.

6b.4.3.3 Conclusion The NRC staff has reviewed SHINEs CSP and various aspects of the SSA and Emergency Plan against the ISG augmenting NUREG-1537, referencing NUREG-1520 and NUREG/CR-6698 as appropriate. Based on its review, the staff has reasonable assurance of the following:

(1) SHINE will have in place a CSP that will be developed, implemented, and maintained to ensure that all nuclear processes are subcritical under normal and all credible abnormal conditions, with an approved margin of subcriticality for safety. Double contingency protection will be provided, where practicable.

(2) SHINE will establish and maintain NCS controls that are subject to facility management measures to support limiting the risk of inadvertent criticality to acceptable limits, as defined in the SSA. Controls will be included in the TS as required by 10 CFR 50.36.

(3) SHINE will have in place a staff of managers, supervisors, engineers, process operators, and other support personnel who are qualified to develop, implement, and maintain the CSP in accordance with the facility organization and administration and management measures.

(4) SHINEs conduct of operations will be based on NCS technical practices that ensure fissile material will be possessed, stored, and used safely consistent with the requirements of 10 CFR Part 50 and certain provisions of 10 CFR Part 70 (5) SHINE will develop, implement, and maintain a criticality accident alarm system in accordance with 10 CFR 70.24 and the facility emergency management program.

(6) An exemption from the requirements of 10 CFR 70.24 for the IU cells and material staging building is warranted.

Based on this review, the NRC staff concludes that the licensees CSP meets the requirements of 10 CFR Part 50 and certain provisions of 10 CFR Part 70, as described by SHINE in SHINE FSAR Section 6b.3, and provides reasonable assurance for the protection of public health and safety, including that of workers and the environment.

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