ML24017A135

Enclosure - Staff Response to the Office of Inspector Generals Special Inquiry Into the U.S. Nuclear Regulatory Commissions Oversight of Research and Test Reactors

ML24017A135
Person / Time
Issue date:	02/02/2024
From:	NRC/NRR/DANU
To:	NRC/OIG
Shared Package
ML24030A502	List: ML24016A252 ML24017A135
References
OEDO-23-00322, OIG I2100162
Download: ML24017A135 (16)

Text

Enclosure Response to the Office of the Inspector Generals Special Inquiry into the U.S. Nuclear Regulatory Commissions Oversight of Research and Test Reactors Executive Summary The U.S. Nuclear Regulatory Commission (NRC) staff conducted a review of the findings presented in the Office of the Inspector Generals (OIG) Case No. I2100162 (Agencywide Documents Access and Management System Accession No. ML23272A039) Special Inquiry into the U.S. Nuclear Regulatory Commissions Oversight of Research and Test Reactors. The OIG inquiry determined that the NRCs research and test reactor (RTR) oversight program failed to identify and address problems with the National Institute of Standards and Technology (NIST) test reactor and other RTRs, as follows:

The NRC failed to identify problems with fuel movement, including precursors to later events.

The NRCs inspection practices often lacked direct observation of activities important to safety.

RTRs other than the NIST reactor experienced significant fuel oversight issues.

The agencys RTR program has not been substantively updated for at least two decades and does not reflect the agencys risk-informed and safety culture positions.

The NRC staff acknowledges the OIGs important role in assuring nuclear reactor safety, its efforts to assess the NRCs oversight program for the Nations RTRs, and the importance of the findings documented in the OIG Special Inquiry report. The NRC staff has also conducted an independent evaluation of its RTR inspection program to identify any areas where improvements in NRC oversight may be implemented. After reviewing the OIG special inquiry report, the staff determined that many of OIGs underlying observations were previously identified by the staff as part of their self-assessment activities. In general, the staffs internal assessment of the NRCs RTR oversight program found no significant gaps; however, the staff did identify several enhancements to internal processes and procedures that could be implemented. The NRC staff made program improvements and identified additional opportunities to enhance the program and update guidance and training in the areas of reactive inspection decision-making process, the level of detail in inspection reports, scheduling onsite inspections, and safety culture, which also correspond to some of the OIGs findings and observations. The staff welcomes the OIG special inquiry report, recognizing that external and internal assessments of NRC programs offer opportunities to gain valuable insights and to enhance the effectiveness and efficiency of our regulatory programs to protect public health and safety.

It is recognized that RTRs are an important component of the Nations nuclear infrastructure, advancing the state-of-the-art research and training for the next generation of nuclear operators and applications. The NRC has ensured the safe operation of RTRs for decades, through the implementation of regulatory licensing and oversight programs that are commensurate with the RTRs associated public health and safety risks. In general, RTRs are inherently safe and pose limited risk to the public and the environment because of their relatively simple designs, low nuclear material inventory, small physical size, and very low power levels as compared to nuclear power reactors. Some RTR reactors, such as the University of New Mexico research reactor is as small as 5 watts, which is equivalent to a light bulb, while the largest (i.e., the NIST

2 reactor) is 20 megawatts; which is substantially smaller than power reactors such as the Ginna Nuclear Power Plant (1775 megawatts) and the most recently constructed light water reactor, Vogtle Electric Generating Plant, Unit 3 (3716 megawatts).

The general requirement for regulation of RTRs is described in Section 104c of the Atomic Energy Act of 1954, as amended (AEA), which inherently acknowledges this lower risk and the important role of RTRs, by requiring the Commission to impose only such minimum amount of regulation as necessary to protect public health and safety, promote common defense and security, and permit the conduct of widespread and diverse research and development. The NRC inspection program for RTRs follows this direction in that the program applies a graded approach to the frequency and scope of inspections commensurate with the risk posed by the facility, to ensure that reasonable assurance of adequate protection of public health and safety and the environment is maintained.

The NRC staff remains committed to the effective licensing and oversight of the Nations RTRs and to protecting public health and safety and the environment. The NRC staffs review of the OIG report identified opportunities to further enhance and improve the RTR inspection program and NRC inspector training, consistent with the improvements identified in the staffs independent self-assessment. The NRC staffs planned actions to enhance the RTR inspection program and procedures are listed in the conclusion section of this response. The NRC staff plans to implement all actions by October 31, 2024. As these actions are implemented and additional insights are gained through future inspection activities, the NRC staff will continue to assess the adequacy of its inspection program and will consider whether additional enhancements may be warranted.

NRC Staff Response to OIG Finding and its Subparts Finding: NRCs inadequate RTR oversight led to a failure to identify and address problems with the NIST test reactor and other RTRs.

Subpart A: NRC Failed to Identify and Address NIST Event Precursors NRC Response to Subpart A:

The NRC staff reviewed subpart A of the report to consider OIGs conclusions in developing enhancements to the RTR oversight program. The OIG conclusions are summarized as follows:

The NRC failed to monitor and address NISTs implementation of its audit committees recommendations.

The NRC did not capture Safety Assessment Committee-identified concerns in its inspection reports.

The NRC failed to identify and address partially latched fuel element issues at NIST on several occasions.

The NRC failed to identify and address fuel movement procedure concerns at the NIST test reactor.

3 The NRC staffs oversight of RTRs is conducted in accordance with the guidance in NRC Inspection Manual Chapter (IMC) 2545, Research and Test Reactor Inspection Program which specifies that the NRCs inspection policies are guided by the AEA. Consistent with IMC 2545, NRC inspections are implemented under the premise that the licensee is responsible for facility safety and compliance with regulatory requirements, and the NRC inspection program is responsible to independently assess the licensees fulfilment of those responsibilities. IMC 2545 further states that the enforcement of these requirements must keep in mind the AEA requirement to impose the minimum amount of regulation [needed]

to protect the health and safety of the public and will permit the conduct of widespread and diverse research and development. Consistent with this approach, IMC 2545 provides further guidance for inspectors that particular attention should be placed on assuring the licensee is not penalized for effectively identifying and correcting its own problems.

The NRC staffs internal self-assessment following the February 2021 NIST event determined that the root causes of the event were subject to regulatory oversight under the existing inspection program guidance and inspection procedures (IPs), and no gaps were identified in the current inspection program. However, the NRC staff acknowledges OIGs underlying observation that the oversight of RTRs could be improved in certain areas and that these improvements would help inform oversight of new projects, such as medical isotope facilities and prospective RTRs based on advanced technology.

The NRC staff ensures reasonable assurance of the safe operation of these facilities through a combination of oversight (routine, supplemental, and reactive inspections) and the facilities licensing basis, taking into consideration the low risk from these facilities. For most research reactors, the licensing basis hypothesizes a very unlikely severe accident to demonstrate that the maximum accidental radiological dose to the public will be less than 100 millirem (mrem)

(i.e., the annual dose limit for individual members of the public resulting from reactor operation, as specified in Title 10 of the Code of Federal Regulations (10 CFR)) 20.1301, Dose limits for individual members of the public,). The NIST event was much less severe than the maximum hypothetical accident considered during initial reactor licensing, which resulted in an acute radiological dose of less than 0.5 mrem, which is significantly lower than the regulatory annual public dose limit of 100 mrem in 10 CFR 20.1301(a) and is a fraction of the U.S. annual average dose of 300 mrem from natural background sources. This demonstrates that the facility was designed and licensed with substantial margin to protect the public.

RTRs are required by technical specifications (TS) to have an independent oversight group, generally referred to as the safety and audit committee, to provide review and audit of the safety aspects of facility operations and provide recommendations to facility management.

The safety committee charter specifies its roles and responsibilities to the licensee, organizational structure, and general reporting requirements. The safety and audit committee issues written reports that may include findings or recommendations to the licensees senior management. While NRC inspectors review issues identified by these groups, they do not typically document such issues in their inspection reports unless the inspector has a related actual or potential safety concern. IMC 0615, Research and Test Reactor Inspection Reports, does not specify documenting findings or recommendations provided by the safety and audit committee in the inspection report. The facility management is responsible for evaluating and dispositioning any recommendations from the licensees safety committee.

However, the inspectors can and do communicate any observations or insights to the licensee during the inspection exit meeting, even if they do not reach the threshold for description in the inspection report. In response to OIGs recommendation, the NRC staff plans to update

4 the guidance in NRC IPs for the review of safety and audit committee audits and findings to include guidance for inspectors on assessing the licensees implementation of the safety and audit committees recommendations, to confirm appropriate follow-up.

NRC licensees are responsible for the safety and security of their facilities and for compliance with regulatory requirements. The NRC staff uses a graded approach to the inspection program for RTRs by applying a frequency and scope of inspections commensurate with the risk posed by the facility to ensure that reasonable assurance of adequate protection of public health and safety and the environment is maintained. Consistent with the NRC inspection policies and guidance, the inspector implements a sampling process when conducting inspections that focuses on reviewing risk-significant activities. These inspections are intended to verify licensee performance and compliance with requirements.

NRC Staff Actions To address OIGs subpart A conclusions and given that the current inspection guidance already contains provisions to directly observe risk significant activities, the NRC staff is placing a greater emphasis on risk significant activities through additional communication with inspectors and orientation of new inspectors on coordination of inspection scheduling with licensees, to increase the opportunities to conduct inspections during reactor operations, fuel movements, and other significant activities. The NRC staff is reemphasizing the need for direct observation of risk significant activities when establishing inspection schedules. The NRC staff has also requested the assistance of the RTR facilities to keep the inspection staff informed of these risk significant activities. Current RTR inspection guidance and policies provide sufficient flexibility to support adding inspection resources on specific inspections as needed to ensure the observation of risk significant activities, such as refueling. Additionally, as discussed above, the NRC staff will update IP guidance to assess the licensees implementation of the safety and audit committees recommendations.

Subpart B: NRCs Inspection Practices Often Lacked Direct Observation of Activities Important to Safety NRC Response to Subpart B:

The NRC staff reviewed subpart B of the report to consider OIGs conclusions in developing enhancements to the RTR oversight program. The OIG conclusions are summarized as follows:

The NRC performed limited direct observations of fuel movements and other licensee activities important to safety.

The NRC did not directly observe fuel element latch verifications in the five years prior to the event.

The OIGs special inquiry report notes that the NRC staffs scheduling of safety inspections was more focused on meeting the required inspection frequency outlined in IMC 2545 instead of focusing on the more risk significant activities occurring at the facility. The NRC staff evaluated the inspection program and determined that it contains sufficient flexibility to optimize the use of inspection resources and conduct inspections commensurate with the safety significance of the RTR. Moreover, the NRC staff schedules inspections considering several factors in addition to the required inspection frequency. This includes the need to conduct pre-inspection planning and post inspection documentation and enforcement activities, the research and development

5 activities occurring at the RTRs, the RTR resources available to support inspection activities, and the inspectors schedules. Although the NRC staff focuses on the NRCs onsite presence along with implementing the inspection frequency in the IMC, the operational activities at RTRs, unlike power reactors, are less predictable and are driven by academic research and other services performed by the facility. They involve frequent changes to RTR operational plans.

These factors make inspection scheduling at RTRs during the conduct of specific activities challenging.

The guidance in IMC 2545 for the performance-based approach to inspection emphasizes observing licensee activities and the evaluating the results of licensee programs, over reviewing procedures and records. However, the NRC staff acknowledges that inspections would benefit from better coordination with operational activities to overcome the inherent challenges in coordinating onsite inspector presence with operational activities that are often rescheduled.

The NRC staffs self-assessment of the RTR oversight program also identified the need to reemphasize existing internal processes and procedures for conducting inspections at RTRs, including an emphasis on observing risk significant operational activities. The staff presented this improvement to the Commission during the Agency Action Review Meeting briefing in June 2023.

NRC Staff Actions Similar to subpart A above, to address OIGs conclusions in subpart B and given that the current inspection guidance already contains provisions to directly observe risk significant activities, the NRC staff is placing a greater emphasis on the coordination of inspection scheduling with licensees to increase the opportunities to conduct inspections during reactor operations, fuel movements, fuel latching, and other risk significant activities. The NRC staff is reemphasizing the importance of direct observation of risk significant activities when establishing inspection schedules and documenting these observations in inspection reports. The NRC staff is also requesting the assistance of the RTR facilities to keep the NRC inspection staff informed of risk significant activities. Current RTR inspection guidance and policies provide sufficient flexibility to support additional inspection resources on specific inspections as needed to ensure the observation of risk significant activities, such as refueling.

Subpart C: NRCs Inadequate Oversight Extends to Other RTRs NRC Response to Subpart C:

The NRC staff reviewed subpart C of the report to consider OIGs conclusions in developing enhancements to the RTR oversight program. The OIG conclusions are summarized as follows:

The NRC did not directly observe fuel movement at Aerotest.

The NRC did not take timely action on damaged fuel at Aerotest. Specifically, the NRC acknowledged the presence of precursors to fuel damage during inspections conducted in 2005, 2007, and 2009.

Aerotest staff made several changes to the facility between 2000 and 2010 without required documentation or revisions to the license or TSs. These changes included:

Use of TRIGA (Training, Research, Isotopes, General Atomics) fuel with different

6 weight percentage and cladding type; and, Operation with a mixed core of fuel elements with different characteristics.

The NRC may have failed to identify the exceedance of an occupational dose limit and the Aerotest facilitys departure from an as low as is reasonably achievable (ALARA) culture.

The NRC failed to identify that Aerotest had TSs that were inadequate under 10 CFR 50.36, Technical specifications.

The NRC had failed to take action on relevant license amendment requests (LARs) the University of Texas (UT) submitted between 2008 and 2012.

The NRC was not on site for UT fuel movements related to the aluminum-cladded elements and missed an opportunity to identify conditions relevant to the licensees noncompliance.

The NRC failed to act on licensee amendment requests from UT to update fuel-related TSs for aluminum-cladded fuel.

The NRC staff evaluated each concern for both Aerotest and UT to identify possible additional enhancements to its oversight of RTRs where appropriate. A summary of the review is provided below.

Aerotest Radiography and Research Reactor (Aerotest)

Evaluation of Actions Taken on Damaged Fuel The TS for TRIGA reactors have specific criteria for identifying fuel elements as damaged.

While swelling in a TRIGA fuel element does not necessarily define the element as damaged fuel, the NRC staff acknowledges that swelling is an early precursor to potential fuel failure. In relation to the 2012 fuel elements cracking event, even though the licensee could not determine the cause of the cracking of the fuel elements, the NRC staff monitored the situation and noted that there was no radiological release to workers or the public as a result of the cracked fuel elements. In January 2012, the NRC staff evaluated the cracking fuel event using Management Directive 8.3, NRC Incident Investigation Program. The NRC staff documented its decision not to conduct a reactive inspection, which was based on the event not resulting in a violation of any TS limit or any release of radioactive material or worker dose. The staff also considered other factors, such as openness, public interest, and public safety, in making this determination.

The NRC staff acknowledges that direct observation of risk significant licensee activities is important to safety, therefore, the NRC staff is reemphasizing the importance of direct observation of risk significant activities when establishing inspection schedules. Additionally, the NRC staff plans to enhance RTR inspector training and updating inspection guidance to increase awareness of precursors that may lead to fuel element damage.

Evaluation of Aerotest Facility Changes The NRC staff evaluated the changes to the cladding type, weight percentage of the fuel, and the operation with a mixed core of fuel elements with different characteristics and determined

7 that those changes occurred in the early 1990-timeframe and the staffs associated oversight activities were adequate. A summary of the NRC staff evaluation is provided below.

Aerotests Hazard Summary Report, dated September 30, 1964 (ML18044A100), specifies the cladding material of the TRIGA fuel elements as aluminum. For initial criticality of the reactor in July 1965, the core contained only aluminum clad TRIGA fuel elements. The TSs approved in 1965, as amended, did not specify a requirement for TRIGA fuel weight percent and cladding type.

The Commissions regulation at 10 CFR 50.59 provides permission for licensees to make changes to their facilities if certain criteria are met. The 50.59 regulation as developed in the 1950s-60s considered if a proposed change involved an unreviewed safety question. The regulation was significantly changed at the end of 1999, but as relevant here, the version in effect in 1992 required the licensee to submit an annual report containing a brief description of any changes, tests, and experiments, including a summary of the safety evaluation of each. In Aerotests annual report Annual Summary of Changes, Tests and Experiments, submitted to the NRC (ML20082C125), for the period ending on June 30, 1991, Aerotest identified adding new TRIGA fuel elements of a different weight percent and cladding material (see figure 1).

This facility change that occurred in 1990 or 1991 would result in Aerotest operating with a mixed core of fuel elements with different percentages and cladding types. Accordingly, and as further discussed below, these changes did not constitute a TS violation. These matters are addressed in the Aerotest annual reports for 1991 and 1992.

Figure 1: Excerpt from Aerotest Annual Report Ending 1991 The following year in 1992, Aerotest reported in its annual report (ML20102B259), other instances of replacing TRIGA fuel elements with new stainless steel fuel elements (see figure 2)

Figure 2: Excerpt from Aerotest Annual Report Ending 1992 The NRC staff inspected the changes to the facility and documented the results in reports dated June 14, 1991 (ML19302E697), and February 23, 1993 (ML20044C419). The inspectors concluded that no changes had been made to the facility or procedures that would require a safety evaluation pursuant to 10 CFR Part 50.59.

Changes and evaluations in accordance with 10 CFR 50.59 have been an area of focus for the NRC staff and the RTR community, to provide clarity and guidance, especially given the RTR

8 licensees infrequent use of this process and the limited guidance that was directly applicable to RTRs. For example, in 2000, the Nuclear Energy Institute (NEI) issued NEI 96-07, Revision 1, Guidelines for 10 CFR 50.59 Implementation, and Regulatory Guide (RG) 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, endorsed the NEI guidance applicable to RTRs. Further, additional RTR-specific 10 CFR 50.59 guidance was issued in February 2022, namely RG 2.8, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments at Non-Power Production or Utilization Facilities, endorsing NEI-21-06, Guidelines for 10 CFR 50.59 Implementation at Non-Power Production or Utilization Facilities (ML21236A089). Additionally, the NRC staff and the RTR community have continued to engage in dialogues on the change process to share experience and gain more clarity. The issuance of these guidance documents as well as the NRC staffs interaction with the RTR community regarding the 10 CFR 50.59 change process have improved the quality of these evaluations.

Evaluation of ALARA Concerns ALARA is a principle for minimizing radiation dose by using the three basic concepts of time, distance, and shielding. The primary ALARA concern for the Aerotest facility was the location of the accounting office adjacent to the waste storage tanks and sump. The NRC staff determined that the maximum radiation levels in the Accounting Office between October 2010 and June 2017 were approximately 0.116 mrem per hour if averaged over the entire quarter. If an individual performed activities for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> a week in that room, that individual would receive an estimated maximum dose of 4.63 mrem per week (or 241 mrem per year). By comparison, the annual limit for occupational workers established in 10 CFR 20.1201(a)(1)(i) is a whole-body total effective dose equivalent (TEDE) of 5 rem. The licensee did implement an ALARA program and, as stated in the Aerotests 2005 safety analysis report (SAR) (ML13120A327), the Aerotest operations policy is that weekly cumulative whole-body exposures are limited to less than 100 mrem, unless higher exposures are specifically approved by the Radiological Safety Officer. An individual working in the accounting office for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> a week would achieve less than 5 percent of the licensees set weekly exposure limit. While the primary responsibility for ALARA and radiation protection lies with the licensee and individual workers, the NRC staff verifies through inspections that the licensee implements certain administrative and engineered controls to support the ALARA principles as defined in 10 CFR 20.1003, Definitions. After Aerotest ceased operation in December 2010, the worker radiation exposure was reduced from approximately 3,000 mrem during 2009 and 2010 to 276 mrem in 2011 and 28 mrem in 2012.

The NRC staff asserts that the implementation of Aerotests ALARA program and the current inspection guidance in IP 69001, Class II Research and Test Reactors, is sufficient to support a conclusion that Aerotest did implement the ALARA principles notwithstanding its prior departure from those principles.

Evaluation of Technical Specifications The Aerotest TS did not contain a fuel temperature safety limit. However, the Aerotest TSs contained other limits, engineered features, and operating conditions to provide reasonable assurance that public and worker safety would be protected. Consistent with the precedent set by other licensees during license renewal, Aerotest recognized the lack of a fuel safety limit in its TSs and, in its license renewal application, proposed TS that included a fuel temperature safety limit. The NRC staff accepted this approach for RTRs to update their TS for license renewal. Aerotests license renewal application was later withdrawn and eventually the facility entered decommissioning. A summary of the NRC staffs evaluation of this issue is provided below.

9 In 1962, the Atomic Energy Commission (AEC) amended 10 CFR 50.36 to specify that TS will be designed to include "those significant design features, operating procedures and operating limitations which are considered important in providing reasonable assurance that the facility will be constructed and operated without undue hazard to the public health and safety." To provide guidance as to matters the AEC generally expected to be covered by TS, Appendix A was added to 10 CFR Part 50 at the same time. In 1965, when the (former) AeroJet operating license was issued, the TS included in the license did not include a fuel temperature limit. In 1968, the AEC revised 10 CFR 50.36 to require an applicant filing of an application for an operating license to propose for Commission review and approval TS derived from the analysis and evaluation presented in the safety analysis report. The 1968 revision required TS for, among other things, safety limits, which are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. Licenses issued under the 1968 revision of 10 CFR 50.36 and those issued under later versions of 50.36 have frequently identified fuel temperature as a safety limit. An example of a fuel temperature safety limit TS is shown below in Figure 4. In 1974, the license was transferred from AeroJet General Corporation to Aerotest Operations, Inc. As part of the license transfer, the NRC approved a revised TS to reflect the licensees name change. Although the 1974 TS did not contain a safety limit for fuel temperature, safe operation was ensured by other TS parameters and limiting conditions for operation including a steady-state power level limit, core configuration, minimum shutdown margin, and maximum reactivity rates to name a few. In 2005, the licensee proposed a TS safety limit in its license renewal application. However, the licensee subsequently withdrew its request for license renewal and permanently shut down in 2010 without incorporating the updated TS. Thus, the proposed revisions to the TS were never implemented.

The NRC staff notes that almost all RTRs have undergone license renewal and that all RTRs, including the few that are currently pursuing license renewal, have adequate safety limits in their TS to help ensure safe operation of the facility.

University of Texas at Austin The NRC staff reviewed the events that led to OIGs conclusion that had the NRC acted on certain LARs submitted between 2008 and 2012, the May 2023 UT event of the UT RTR using aluminum clad fuel elements may have been averted. While the NRC staff determined that additional license amendments would have been required for the facility to operate with aluminum clad fuel elements, UT had not requested any licensing action regarding use of these elements. In terms of licensing action timeliness, the NRC staff has since implemented several initiatives including licensing review metrics and the use of acceptance reviews to ensure actions are processed in a timely manner. A summary of the NRC staffs review for this OIG conclusion is provided below.

The UT LAR dated March 22, 2004 (ML040910231), requested a change to paragraph 2.B.(2) of license R-129 to increase the special nuclear material (SNM) limit to receive, possess, and use up to 9.5 kilograms of contained uranium-235 in the form of TRIGA fuel to accommodate a licensee delay in shipping out spent fuel, to receive a replacement instrument fuel element, and to proceed with the planned acquirement of additional TRIGA fuel by the end of 2005.

This LAR did not seek authority to use aluminum-clad fuel, nor did any other subsequent LAR request such authorization. The current UT license, appendix A, TS 5.3.1, Fuel Elements, item c, only authorizes a stainless-steel design feature for TRIGA elements (see figure 3).

10 Figure 3: Excerpt from University of Texas 2004 Technical Specifications For the licensee to use aluminum clad TRIGA fuel elements in the reactor core, a change to TS 5.3.1(c) would be required. Also, aluminum clad TRIGA fuel elements have a lower safety limit compared to the stainless-steel fuel elements (due to a lower melting temperature) to ensure the integrity of the cladding is maintained. Therefore, a change to TS 2.1, Safety Limit, would also be required for the use of aluminum clad TRIGA fuel elements. In the 2004 LAR, the licensee only requested a special nuclear material possession limit increase. It is common practice for RTRs to receive and possess fuel while they develop the technical justification and receive the necessary regulatory approvals to use the fuel in the core. The NRC must authorize the revised possession limit before the receipt of the fuel. The licensee is responsible for ensuring that a different type of fuel is adequately controlled and not used in the core until it is authorized by the license. The licensee is also responsible for fully describing the changes when requesting to amend the license, which includes TS in accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Currently, the NRC has authorized only Dow Chemical Company, University of Utah, and U.S. Geological Survey research reactors to use a mixed reactor core of stainless-steel and aluminum clad TRIGA fuel elements. For example, the TS safety limit for a facility that uses a mix of aluminum and stainless-steel clad TRIGA elements would be as follows (see figure 4).

Figure 4: Excerpt from University of Utah Technical Specifications - Safety Limit Both the 2008 and 2010 UT LARs (ML080920755 and ML101241147, respectively) proposed a change to a TS definition and only requested to change the definition of Fuel Element, Standard to include aluminum clad fuel elements similar to a definition used by General Atomics. Definitions only provide uniform interpretation of terms that are used in the TS. In the 2008 and 2010 LARs, the licensee did not specifically request to use aluminum clad TRIGA fuel elements in the core. If it had wanted to use aluminum clad TRIGA fuel, the licensee was required to develop a technical justification to support the determination that aluminum clad

11 TRIGA fuel elements are safe to use in the core (see figure 4 above). In addition, UT would have needed to request a change to the TRIGA fuel element cladding design feature, as specified in TS 5.3.1.c, which requires the use of stainless-steel cladding (see figure 3, above).

TS 5.3.1 requires that TRIGA fuel elements for use in the UT core are clad with 304 stainless-steel. The NRC staff also notes that the fuel element cladding thickness differs between stainless-steel and aluminum clad fuel elements. The licensee did not provide a technical justification or propose a change to TS 5.3.1 in either the 2008 or the 2010 LAR submitted to the NRC.

On February 8, 2012 (ML12082A145), UT submitted an LAR for another change to the definition of Fuel Element, Standard that only references stainless-steel clad fuel elements and not aluminum clad TRIGA fuel elements (see figure 5). UTs basis for the change was editorial in nature and did not affect safety. On August 19, 2016 (ML16252A219), UT requested that the NRC include these proposed changes to the facility operating license for license renewal.

Hence, in its licensing action requests between 2008 and 2016, UT did not specifically request to use aluminum clad TRIGA fuel elements in the core and did not develop a technical justification to support the determination that aluminum clad TRIGA fuel elements are safe to use in the core. As such, the NRC did not review, nor approve, the use of aluminum clad TRIGA fuel elements in the UT core.

Figure 5: Excerpt from University of Texas 2012 license amendment request The timeliness of licensing actions is a focus area for the NRC staff and the RTR community, and the NRC staff has implemented several initiatives including licensing guidance, robust pre-application engagement, and timely acceptance reviews to ensure timely actions on all LARs. That said, the NRC staff concludes that more rapid review and/or issuance of any of the 2004, 2008, and 2010 LARs would not have prevented the May 2023 event because the requests were unrelated to authorization to use aluminum clad TRIGA fuel in the core.

NRC Staff Actions Regardless of the assessment of the LARs submitted from 2008 to 2012, to address OIGs conclusions in subpart C, the NRC staff plans to enhance inspector training and guidance on the precursors that could lead to fuel element damage. Additionally, as previously described, the NRC staff is placing additional emphasis on observing risk-significant operational activities during inspections. In addition, the NRC staff has made many updates to its processes and procedures (e.g., 10 CFR 50.59 guidance) that have resulted in improvements that address OIGs additional conclusions.

12 Subpart D: RTR Inspection Program Policy and Guidance are Outdated NRC Response to Subpart D:

The NRC staff reviewed Subpart D of the report to consider OIGs conclusions in developing enhancements to the RTR oversight program. These conclusions are summarized as follows:

The NRC has not implemented the risk-informed approach recommendations from NUREG-2150, A Proposed Risk Management Regulatory Framework in the RTR inspection program policy and guidance.

Safety culture was not implemented in the RTR inspection program as recommended in the NRCs Safety Culture Policy Statement.

The NRC has not updated the safety aspect of IMC 2545, Research and Test Reactor Inspection Program, since 2004, which is inconsistent with IMC 2545.

The last major revision to the safety inspection program documents was in 2004, and the 2004 IPs underestimate the resources needed to complete all requirements.

The RTR inspection program does not have a self-assessment process to determine if the program meets its established goals and intended outcomes.

Lack of a risk-informed approach in RTR inspection program At the request of the NRC Chairman, an NRC Risk Management Task Force was chartered to develop a strategic vision and options for adopting a more comprehensive, holistic, risk-informed, performance-based regulatory approach for reactors (including RTRs), materials, waste, fuel cycle, and transportation that would continue to ensure the safe and secure use of nuclear material. The task force issued NUREG-2150 in April 2012, which describes a proposed risk management regulatory approach that could be used to improve consistency among the NRCs various programs and discusses implementing such a framework for specific program areas. NUREG-2150, in part, describes the implementation of a proposed risk management regulatory framework for nonpower reactors. The task force developed findings and recommendations on changes that would be needed to ensure that the proposed risk management framework would be implemented in 10 to 15 years. In 2012, the NRC Chairman requested the staff to review NUREG-2150 and provide options and recommendations to the Commission, including the potential for adopting the proposed risk management regulatory framework through a Commission policy statement (ML121660102).

On December 18, 2015, the NRC staff summitted SECY-15-0168, Recommendations on Issues Related to Implementation of a Risk Management Regulatory Framework, dated December 18, 2015 (ML15265A488), in response to the Chairmans request. The NRC staff recommended that the Commission not develop and issue an agencywide risk management policy statement. The NRC staff based its recommendation on an analysis of the expected benefit of a policy statement compared to the resource expenditure to create the statement and was informed by public feedback. The NRC staff compared the resource expenditure to create and implement a policy statement across the program offices and concluded that NRC

13 resources were not justified and were better focused on issues of greater safety significance.

On March 9, 2016, the Commission issued the related staff requirements memorandum (ML16069A370) and approved the NRC staffs recommendation to refrain from developing an overarching, agencywide risk management policy statement. As a result of the Commissions direction, the NRC staff did not implement the recommendations of NUREG-2150 in the RTR inspection program policy and guidance.

The NRC staffs position is that the current oversight and inspection program for RTRs is risk informed based on the risk posed to the public by facility operation. The RTR inspection program uses a 2 megawatts threshold to define the inspection program and the frequency by which it is implemented at a facility. NRC regulations do not require RTRs to develop and maintain a probabilistic risk assessment for their facilities. A probabilistic risk assessment-oriented approach for RTRs would be resource intensive to develop and maintain and would provide limited benefit in informing facility risk assessment. Instead, consistent with existing NRC guidance, during licensing as described in the safety analysis reports for RTRs, licensees perform deterministic accident analyses that typically demonstrate a large margin to safety limits. The NRC staffs position is that the current RTR licensing and oversight approach is appropriately risk-informed; in addition, the NRC staff is reemphasizing the importance of direct observation of risk significant activities when establishing inspection schedules.

In spring 2020, the NRC staff developed and piloted supplemental guidance to augment the decision-making guidance in Management Directive 8.3 aimed at enhancing the NRCs oversight response to an event at an RTR. As a result of the February 2021 NIST event, the NRC staff identified further enhancements to the pilot supplemental guidance. These enhancements were included in the supplemental guidance, which was issued in October 2022 (ML22257A162).

Lack of safety culture element in RTR inspection program Operating experience demonstrates a clear nexus between safety culture and the occurrence of events at a facility. The NRC staff acknowledges the importance of positive safety culture traits at RTR facilities within the NRC oversight program. The current qualification program for RTR inspectors includes training on the Commissions Safety Culture Policy Statement, safety culture traits, and safety conscious work environment. In addition, the NRC staff plans to include safety culture guidance in the RTR inspection program, procedures, and training to provide the inspectors additional tools to detect potential weaknesses in a licensees safety culture.

Moreover, the NRC staff is increasing its focus on sharing operational experience within the RTR community. For example, on June 13, 2023, the NRC staff issued Information Notice 2023-03, Recent Human Performance Issues at NonPower Production and Utilization Facilities (ML23059A539), outlining operational issues for several facilities that highlighted the need for a strong safety culture. Also, the NRC staff is focusing on communicating operating experience insights and lessons learned in appropriate venues, including trade conferences and periodic meetings with the RTR community.

To communicate the importance of safety culture and operating experience, the NRC staff developed a public-facing nonpower production and utilization facilities (NPUFs) inspection report webpage and is developing an operational experience dashboard that is publicly accessible to enhance sharing of reactor operating experience in the RTR community.

Additionally, the NRC staff aided RTR licensees in accessing the International Atomic Energy Agencys (IAEAs) international operating experience database, known as the Incident

14 Reporting System for Research Reactors, to enable the sharing of international operational experiences. During public meetings and conferences, the NRC staff will continue to promote the use of the IAEA operating experience database to RTR licensees to share operating experience and lessons learned that could help prevent the occurrence of similar events at their respective RTRs.

Lack of safety inspection program updates The NPUF oversight and licensing branches in the Office of Nuclear Reactor Regulation hold an annual joint branch meeting to review the licensing and oversight activities at each RTR facility over the prior year to identify common performance issues or concerns, operational experience, and trends for consideration in current regulatory programs including any recommended changes, if needed. Additionally, the NRC staff made several enhancements to and reviewed the RTR inspection program since 2020. These enhancements and reviews are documented in an internal, non-public self-assessment report. As indicated above, the NRC staffs recent internal self-assessment of the NRCs RTR oversight program found no significant gaps in the inspection program and guidance; however, several possible enhancements to internal processes and procedures were identified. The NRC staff has determined that the combination of site-specific licensing bases, a robust emergency response infrastructure, and the current NRC oversight program are appropriately risk-informed by the application of inspection scope and frequency and continue to provide appropriate protection to public health and safety and the environment for RTRs. Further, to maximize the benefit of its self-assessment, the NRC staff plans to further enhance its current program assessment activities to formalize a recurring self-assessment process and document the conclusions reached, to assure that the oversight program continues to meet its established goals of protecting public health and safety. These results will be documented and presented to NRC management. The NRC staff also plans to incorporate guidance in IMC 2545 on performing a periodic review of the RTR IMCs and IPs.

Inadequate resource estimates to support inspection requirements In terms of the overall RTR oversight program resources, NRC management continually evaluates staffing resource needs and skills gaps in the budgeting and strategic workforce planning processes. These activities recently identified the need for additional inspection resources due to the growth in the number of future NPUFs. Additionally, NRC management has used flexibilities to provide better balance of inspector responsibilities that considers the complexity of the RTR facilities and has encouraged and implemented cross-training and cross-qualifying staff to provide flexibility in assigning branch resources to inspections. The combination of these processes and managements use of existing flexibilities for assigning resources ensure that inspection resources are available to support inspections at RTR facilities.

Regarding resources estimated for specific inspection activities, as stated in the IPs, the resource estimate for each IP is provided for planning purposes, and an inspector can exceed the direct inspection resource estimate, if needed, to complete an inspection activity. The NRC staff reemphasized the flexibility in the inspection resource estimates with the inspectors to ensure their awareness. In addition, the NRC staff plans to formalize a recurring self-assessment process that includes evaluating estimated hours and sample sizes for RTR IPs to support adequate inspection resources. If new or revised information is presented that could impact inspection resources, the NRC staff will evaluate and revise, as necessary, the estimated hours in the IPs using the process to be described in IMC 2545 for updating RTR IMCs and IPs.

15 NRC Staff Actions In summary, to address the conclusions described in subpart D of the OIG report, the NRC staff is reemphasizing the direct observation of risk significant activities when establishing inspection schedules. The NRC staff plans on enhancing inspector training and IPs to increase awareness and recognition of potential weaknesses in a licensees safety culture. Additionally, the NRC staff will incorporate guidance in IMC 2545 on performing a periodic review of the RTR IMCs and IPs. During the periodic review of the IPs, the NRC staff will evaluate individual IP hour estimates to ensure the sufficiency of resources. Additionally, the NRC staff will continue to share safety culture operating experience with RTR licensees in different venues, such as during public meetings, at conferences, and on the NRC public website.

Conclusion The NRC staff appreciates OIG's efforts and its valuable insights concerning the NRCs oversight program for the Nations RTRs. The NRC staff has determined that the combination of site-specific licensing bases, a robust emergency response infrastructure, and the current NRC oversight program is appropriately risk-informed and continues to provide adequate protection of public health and safety and the environment for RTRs. The NRC staffs recent internal, non-public self-assessment, of the NRCs RTR oversight program found no significant gaps that would impact the safe operation of RTRs. The self-assessment identified several enhancements to internal processes and procedures, some of which OIG also identified in its report. As a result, the NRC staff is taking the following actions:

Reemphasize the importance of direct observation of risk significant activities when establishing inspection schedules at RTRs.

Update inspection guidance on assessing the licensees implementation of the safety and audit committees recommendations for appropriate follow-up.

Enhance inspector training and inspection procedures to increase awareness and recognition of potential weaknesses in a licensees safety culture.

Enhance inspector training and guidance on the precursors that could lead to fuel element damage.

Develop a public NPUFs inspection report webpage and an operational experience dashboard that is easily accessible for the licensees and members of the public to enhance sharing of reactor operating experience within the RTR community and continue to communicate the importance of using the IAEAs operating experience database for RTRs in sharing operational experience and lessons learned.

Enhance the current annual licensing and oversight program review to include a formalized self-assessment process and document the assessment conclusions to determine whether the RTR oversight program continues to meet its established goals of protecting public health and safety and the environment.

16 Incorporate guidance in IMC 2545 for performing a periodic review of IMCs and IPs for RTRs.

Date of planned completion for all actions: October 31, 2024 Point of

Contact:

Travis Tate, NRR/DANU/UNPO