ML23263A976

Regents of the Ucd Issuance of Amendment No. 10 to Renewed Facility Operating License R-130

ML23263A976
Person / Time
Site:	University of California-Davis
Issue date:	09/22/2023
From:	Linh Tran NRC/NRR/DANU/UNPL
To:	Gibeling J McClellan Nuclear Research Center
References
EPID L-2023-NFA-0009
Download: ML23263A976 (1)

Text

Dr. Jeffery C. Gibeling Interim Vice Chancellor for Research Department of Computer Science University of California Davis, CA 95616

SUBJECT:

REGENTS OF THE UNIVERSITY OF CALIFORNIA - ISSUANCE OF AMENDMENT NO. 10 TO RENEWED FACILITY OPERATING LICENSE NO. R-130 TO AMEND THE TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA-DAVIS MCCLELLAN NUCLEAR RESEARCH CENTER (EPID NO. L-2023-NFA-0009)

Dear Dr. Gibeling:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 10 to Renewed Facility Operating License No. R130 for the University of California-Davis (UCD)

McClellan Nuclear Research Center (MNRC) Training, Research, Isotope, General Atomics research reactor. The amendment consists of changes to the renewed facility operating license and technical specifications (TSs) in response to your application dated September 15, 2023 (Agencywide Documents Access and Management System Accession No. ML23258A242), as supplemented by letter dated September 19, 2023 (ML23262B520).

The amendment revises UCD MNRC TS 4.1.4, Fuel Parameters, and adds a specification to TS 5.3.1, Reactor Core, to provide the licensee operational flexibility to reconfigure the reactor core.

September 22, 2023

J. Gibeling 2

A copy of the related safety evaluation is also enclosed. If you have any questions, please contact me at 301415-4103, or by email at Linh.Tran@nrc.gov.

Sincerely, Linh N. Tran, Senior Project Manager Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Docket No. 50607 License No. R130

Enclosures:

1. Amendment No. 10 to Renewed Facility Operating License No. R130

2. Safety Evaluation cc w/enclosures: GovDelivery Subscribers Signed by Tran, Linh on 09/22/23

ML23263A976 NRR058 OFFICE NRR/DANU/UNPL/PM NRR/DANU/UNPL/LA OGC NAME LTran NParker JWachutka DATE 9/20/2023 9/21/2023 9/21/2023 OFFICE NRR/DANU/UNPL/BC NRR/DANU/UNPL/PM NAME JBorromeo (HCruz for)

LTran DATE 9/21/2023 9/22/2023 REGENTS OF THE UNIVERSITY OF CALIFORNIA DOCKET NO. 50607 UNIVERSITY OF CALIFORNIA-DAVIS MCCLELLAN NUCLEAR RESEARCH CENTER AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 10 License No. R130 1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to Renewed Facility Operating License No. R-130, filed by the Regents of the University of California (the licensee) on September 15, 2023, as supplemented on September 19, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E.

The issuance of this amendment is in accordance with 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, of the Commissions regulations and all applicable requirements have been satisfied; and F.

Prior notice of this amendment was not required by 10 CFR 2.105, Notice of proposed action, and publication of a notice for this amendment is not required by 10 CFR 2.106, Notice of issuance.

2.

Accordingly, the license is amended as described in Attachment 1 to this license amendment and by changes to the Technical Specifications as indicated in. Paragraph 2.C.2 of Renewed Facility Operating License No. R-130 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No. 10, are hereby incorporated in their entirety in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION for Joshua M. Borromeo, Chief Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Attachments:

1. Changes to Renewed Facility Operating License No. R130

2. Changes to Appendix A, Technical Specifications Date of Issuance: September 22, 2023 Holly D.

Cruz Digitally signed by Holly D. Cruz ATTACHMENT 1 TO LICENSE AMENDMENT NO. 10 RENEWED FACILITY OPERATING LICENSE NO. R130 DOCKET NO. 50607 Replace the following page of Renewed Facility Operating License No. R130 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Renewed Facility Operating License No. R-130 REMOVE INSERT Amendment No. 10 September 22, 2023 C. This license shall be deemed to contain and is subject to the conditions specified in 10 CFR Parts 20, 30, 50, 51, 55, 70, and 73 of the Commissions regulations; is subject to all applicable provisions of the Act; and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

1.

Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 1,000 kilowatts (thermal) in accordance with the limitations in the Technical Specifications.

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No. 10, are hereby incorporated in their entirety in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Physical Security Plan The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The approved plan, entitled Physical Security Plan for the U.C. Davis/McClellan Nuclear Research Center, dated January 11, 2022, consists of documents withheld from public disclosure pursuant to 10 CFR 73.21.

D. This license is effective as of the date of issuance and shall expire at midnight, 20 years from its date of Issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Andrea D. Veil, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A, Technical Specifications Date of Issuance: November 21, 2022 ATTACHMENT 2 TO LICENSE AMENDMENT NO. 10 RENEWED FACILITY OPERATING LICENSE NO. R130 DOCKET NO. 50607 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Technical Specifications REMOVE INSERT 18 18 24 24

18 Amendment No. 10 September 22, 2023 Specification-1.

A daily check of the core shall be made to verify only stainless steel clad 20/20 and 30/20 elements are only located Hex Rings C through G.

2.

Prior to removal of any fuel element it shall be verified that the core is subcritical by more than the calculated worth of the most reactive fuel element being moved.

3.

Prior to manual removal of any control rod it shall be verified that the core is subcritical by at least $0.50 with the highest worth control rod in the full-out position.

Basis

1. This specification provide verifications that core configuration will not deviate from the core configuration analyzed for in the SAR.

2. and 3. These specifications ensure the core will remain shut down during fuel and control rod movements and inspections.

4.1.4 Fuel Parameters Applicability - This specification applies to the surveillance requirements for the fuel elements.

Objective - The objective is to verify the continuing integrity of the fuel element cladding.

Specification -

1.

All fuel elements shall be inspected for damage or deterioration and measured for length and transverse bend at least at quinquennial intervals.

2.

An analysis of any irradiation facility installed in the central cavity of this core shall be done before it is used with this core.

3.

No single element shall be operated at a power level above 17.69 kW (as analyzed).

Basis-

1. The above specifications assure that the fuel elements shall be inspected regularly and the integrity of the lead fuel elements shall be maintained.

2. and 3. Will provide assurance that the thermal hydraulic analysis provided in the SAR is always bounding.

4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Applicability - This specification applies to the surveillance of the control rods.

Objective - The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rods.

24 Amendment No. 10 September 22, 2023 positions occupied by in-core experiments, irradiation facilities, graphite dummies, control rods, and startup sources.

3. The reflector, excluding experiments and irradiation facilities, shall be graphite. A reflector is not required if the core has been defueled.

4.

G ring core locations may be empty (water-filled).

5.3.2 Reactor Fuel Applicability - These specifications apply to the fuel elements used in the reactor core.

Objective - The objective is to assure that the fuel elements are of such design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specification - The individual unirradiated TRIGA fuel elements shall have the following characteristics:

1.

Uranium content: 20 or 30 wt % uranium enriched nominally to less than 20% U-235.

2.

Hydrogen to zirconium atom ratio (in the ZrHx): 1.60 to 1.70 (1.65+/- 0.05).

3.

Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.

5.3.3 Control Rods and Control Rod Drives Applicability - This specification applies to the control rods and control rod drives used in the reactor core.

Objective - The objective is to assure the control rods and control rod drives are of such a design as to permit their use with a high degree of reliability with respect to their physical, nuclear, and mechanical characteristics.

Specification -

1.

All control rods shall have scram capability and contain a neutron poison such as stainless steel, borated graphite, B4C powder, or boron and its compounds in solid form. The shim and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air filled follower and be sealed in an aluminum tube.

2.

The control rod drives shall be the standard GA rack and pinion type with an electromagnet and armature attached.

5.4 Fissionable Material Storage Applicability - This specification applies to the storage of reactor fuel at a time when it is not in the reactor core.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 10 TO RENEWED FACILITY OPERATING LICENSE NO. R130 REGENTS OF THE UNIVERSITY OF CALIFORNIA UNIVERSITY OF CALIFORNIA-DAVIS MCCLELLAN NUCLEAR RESEARCH CENTER DOCKET NO. 50607

1.0 INTRODUCTION

By letter dated September 15, 2023 (Agencywide Documents Access and Management System Accession No. ML23258A242), as supplemented by letter dated September 19, 2023 (ML23262B520), the Regents of the University of California (the licensee) submitted a license amendment request (LAR) to amend the Appendix A, Technical Specifications, to Renewed Facility Operating License No. R130 for the University of California-Davis (UCD) McClellan Nuclear Research Center (MNRC) Training, Research, Isotope, General Atomics (TRIGA) research reactor. Specifically, the licensee proposed to revise technical specification (TS) 4.1.4, Fuel Parameters, and add a new specification to TS 5.3.1, Reactor Core, to provide the licensee operational flexibility to reconfigure the reactor core.

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC, the Commission) staff reviewed the licensees LAR and evaluated the proposed TS changes based on the following regulations and guidance:

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Section 50.36, Technical specifications, which provides the requirements for TSs to be included in facility operating licenses, including research reactor licenses. Section 50.36(c)(3), Surveillance requirements, of 10 CFR states, in part, that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Section 50.36(c)(4), Design features, of 10 CFR states, in part, that design features to be included in TSs are those features of the facility, which, if altered or modified, would have a significant effect on safety and are not covered in the TSs as safety limits, limiting conditions for operation, or surveillance requirements.

Part 20, Standards for Protection against Radiation, of 10 CFR, which establishes the regulatory requirements for protection against ionizing radiation resulting from activities conducted under licenses issued by the NRC.

NUREG1537, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content, Chapter 14, Technical Specifications, Appendix 14.1, Format and Content of Technical Specifications for Non-Power Reactors (ML042430055), which provides guidance to licensees preparing research reactor applications and TSs.

NUREG1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, Chapter 14, Technical Specification (ML042430048), which provides guidance to the NRC staff for performing reviews of proposed TSs.

3.0 TECHNICAL EVALUATION

3.1 Background

In its LAR, as supplemented, the licensee indicated that during the facilitys 2023 fuel inspection, one TRIGA fuel element showed inward pitting and unusual discoloration in the cladding, which did not meet the requirements in TS 3.1.4, Fuel Parameters, Specification 4, which states that a fuel element shall be considered damaged and must be removed from the core if visual inspection finds bulges, gross pitting, or corrosion. The licensee reported this issue to the NRC in Event Notification 56643, dated July 28, 2023 (https://www.nrc.gov/reading-rm/doc-collections/event-status/event/2023/20230731en.html#en56643).

As a result of this finding of a failed fuel element, the licensee inspected all the adjacent fuel elements and all other fuel elements that had not been inspected within the last year (approximately 85 percent of all in-core fuel elements). During this inspection, the licensee observed several fuel elements in G ring (the outer-most fuel ring of the reactor core) that were difficult to remove from the grid plate, with two G ring fuel elements being very difficult to remove from the core. In one of these G ring positions, the licensee observed a small metal protrusion just below the upper grid plate that appeared to prevent the reinsertion of a fuel element or a graphite element (which has a slightly smaller diameter) in this position. The licensees review determined that operating the reactor with an empty (water-filled) position where the protrusion was discovered was prohibited by the requirement in TS 5.3.1, Reactor Core. Since neither a fuel element nor a graphite element could be inserted in this position, the licensee was prevented from restarting the reactor by this TS.

The licensees investigation concluded that the most likely cause of the issues observed in the G ring locations was that a design/construction flaw was introduced during the power upgrade that was performed in the mid-1990s to increase the reactor steady-state power from the initial licensed power level of 1.0 megawatt (MW) to 2.0 MW. During this upgrade, the core was defueled and the upper and lower grid plates were removed in order to replace the original radial grid plates with new hexagonal grid plates. This included adding six shim plates between the G ring and the core barrel. The purpose of these shim plates was to reduce the flow area outside of the fuel, and thus the amount of water that could bypass the inner part of the core. Depictions of the shim plates are shown in turquoise in the figure below.

Figure 1 - Shim Plates (in turquoise)

One remnant of the original core that could not be removed during the upgrade was the upper core ring, which contained 6 tabs that protruded towards the core. The licensees review of records from the upgrade project determined that these tabs should have been machined (underwater) flush with the rest of the ring. The licensee stated that this task may have proved too difficult and did not appear to have been done properly. As a result, the licensee concluded that these tabs most likely forced the shim plates towards the core and narrowed the coolant channel gap between the shim plates and the fuel elements located in G ring.

As part of its resolution of this issue, the licensee designed and built a measurement tool to identify which G ring positions had at least the coolant channel gap width (0.065 inches) used in the thermal-hydraulic (T-H) calculations found in section 4.7.1.5, Hot Channel (103), of chapter 4.0, UCD/MNRC TRIGA Reactor, of the MNRCs license renewal application updated safety analysis report (SAR), dated July 6, 2020 (ML20238B984). The results of this inspection showed that 4 out of the 6 shim plates were out of place and that none of the grid positions along those shim plates had the required coolant channel gap width. The shim plates could not be moved or shifted when modest force was applied to them by the inspection tool. The licensee concluded that the position of the 6 shim plates was likely static and had not changed significantly since the upgrade project that had been implemented approximately 25 years ago.

As a result, the licensee determined that 21 of the 30 G ring positions had insufficient clearance between the shim plates and the fuel elements and could not be used for fuel elements, thus, requiring a core reconfiguration and TS amendment.

Prior to the 2023 fuel inspection, the licensees fuel inspection TS requirements only required inspections of the lead elements, which were defined as the fuel elements in the C ring and those surrounding the transient control rod. With the issuance of the renewed license, dated November 21, 2022, the licensees TSs incorporate the fuel inspection guidance provided in NUREG-1537, Part 1, chapter 14, appendix 14.1, section 4.1, Reactor Core Parameters, item (6), Fuel Parameters, which states, in part, that all fuel elements should be inspected on at least a 5-year cycle. As a result, fuel elements in the G ring positions probably remained static and unobserved for many years prior to the 2023 fuel inspection.

In addition to the LAR, the licensee provided additional information during a public meeting conducted on August 30, 2023. The meeting notice, the licensees presentation, and the NRC staffs summary of the meeting are available at ML23256A372.

The NRC staff reviewed the licensees T-H analysis as part of its license renewal review and documented its findings in the safety evaluation report (SER) issued with the renewed license, Safety Evaluation Report, Renewal of the Facility Operating License for the University of California-Davis McClellan Nuclear Research Center TRIGA Research Reactor, November 21, 2022 (ML22214B831). The SER concluded that the licensees T-H analysis, and the fuel element power level limit of 17.69 kilowatt (kW) at a steady-state power level of 1.0 MW, as required by TS 4.1.4, Specification 3, was acceptable.

The NRC staff review of the proposed TS changes focused on ensuring that the licensee had the flexibility to reconfigure the reactor core to support continued operation while maintaining the existing approved T-H analysis described above.

3.2 Proposed Changes to TS 4.1.4, Specification 3 As stated in its LAR, the licensee proposed the following changes to TS 4.1.4, Specification 3, which are denoted below using bold to indicate additional text and strikeout to indicate deleted text.

Current TS 4.1.4, Specification 3, states:

3. No single element may be operated at a power level above 17.69 kW (as analyzed) at a steady state power level of 1.0 MW.

Proposed TS 4.1.4, Specification 3, states:

3.

No single element may shall be operated at a power level above 17.69 kW (as analyzed) at a steady state power level of 1.0 MW.

As stated in its LAR, the licensee plans to reconfigure the reactor core such that the G ring locations with insufficient clearance are filled with graphite elements. In order to have sufficient excess reactivity, the licensee indicated that it would need to move some higher reactivity fuel elements, e.g., 30 uranium (U) weight percent with 20 percent U-235 enrichment (30/20), to the inner-most fuel rings of the core (C and D ring). The licensee indicated that this proposed core configuration is very similar to the limiting core configuration previously analyzed in chapter 4.0 of the SAR and is illustrated in the figure below.

Figure 2 - Proposed Core Configuration The licensees analysis of the proposed core configuration determined that the overall effective hot channel peaking factor is less than that of the limiting core configuration (beginning of life),

which has the highest peaking factors of any analyzed core. As the proposed core ages and some burnup is introduced in the new 30/20 elements located in the C ring, the peaking factors remain essentially unchanged (though the peaking factors do very slightly decrease with burnup). The analysis also determined that the hot channel thermal power (hottest single element) was 17.80 kW at 1.0 MW, which exceeds the limit of 17.69 kW at 1.0 MW as stated in current TS 4.1.4, Specification 3, and, thus, the proposed core configuration would not be allowed unless TS 4.1.4, Specification 3 were to be amended to remove the 1.0 MW criterion.

The licensee indicated that the proposed change to TS 4.1.4, Specification 3 to remove the 1.0 MW criterion is necessary to allow reactor operation using the proposed core configuration. The licensee stated that a reduction in steady-state power level to less than 1.0 MW is needed to limit the hottest element to a power level of 17.69 kW. For the proposed core configuration, the steady-state power level would be limited to 993 kW to ensure that the hottest element power level does not exceed the previously approved TS limit of 17.69 kW. The licensee indicated that the required SCRAM setpoints would also need to be adjusted and that the maximum steady-state power level would need to be administratively limited to help ensure compliance with the proposed changes to TS 4.1.4, Specification 3.

The licensee indicated that, based on the results of analysis of the hottest fuel element power level, it would use the 10 CFR 50.59, Changes, tests, and experiments, process to adjust (lower) the required SCRAM setpoints and administratively limit maximum steady-state power level accordingly, to ensure that operation of the hottest fuel element power level remains within the TS limit of 17.69 kW. The licensee indicated that an analysis of the hottest fuel element power level will be performed annually, after any significant core change, or in support of any experiment to be placed in the central cavity of the core. In its supplemental letter dated September 19, 2023, the licensee stated that the performance of the analysis demonstrating that the hottest single fuel element power level remains below TS 4.1.4, Specification 3 is provided in the licensees Fuel Document, UCD MNRC-0011-OMM.

Finally, the licensee indicated that due to excess reactivity (fuel elements) limitations, the licensee administratively lowered the maximum steady-state power level to 800 kW almost a year ago in October of 2022, and has no plans of operating the reactor at power levels above 800 kW.

NRC Staff Review The NRC staff review finds that the proposed deletion of the reference to a steady-state power level of 1.0 MW in TS 4.1.4, Specification 3 is necessary to provide the licensee operational flexibility to reconfigure the reactor core given the loss of 21 G ring locations for fuel elements.

The proposed TS 4.1.4, Specification 3 would maintain the power level limit of 17.69 kW in the hottest single fuel element, which would preserve the results of the T-H analysis described in the SAR that had been previously approved by the NRC.

The NRC staff also finds that the requirement to analyze the hottest fuel element power level as described in TS 4.1.4, Specification 3 would be supported by the licensees plan, as described in the LAR, to perform an analysis of the hottest single element power level annually, after any significant core change, or in support of any experiment to be placed in the central cavity of the core, which is sufficient to ensure that the facility operates within the assumptions of its previously approved T-H analysis. In addition, the NRC staff finds the licensees plan to incorporate the requirements to perform the analysis of the hottest single element power level in the licensees Fuel Document, UCD MNRC-011-OMM, acceptable.

The proposed core configuration would eliminate 21 G ring locations for fuel elements by replacing them with 20 graphite elements and one empty (water-filled) position. The NRC staff review finds that replacing 21 of 30 G ring locations with graphite elements (which tend to act as neutron reflectors) and one empty (water-filled) position (which tends to act as a neutron moderator), would slightly shift power toward the center of the reactor core (C and D rings), and potentially result in a slight increase in some fuel element power levels. The NRC staff review finds that the licensees LAR analysis shows little change in the peaking factors, which is consistent with the licensees reported maximum power of 17.80 kW at 1.0 MW. The NRC staff review finds that the licensees analysis results are reasonable given that the fuel elements in G ring locations typically operate at a much lower power level than the inner rings simply due to the loss of neutrons to the core periphery.

The NRC staff review also finds that the licensees indicated plan to operate at a reduced steady-state power level (800 kW) would be effective at ensuring that the maximum fuel element power level remains well below the TS limit of 17.69 kW.

Based on the above, the NRC staff concludes that the licensees proposed change to TS 4.1.4, Specification 3 to change may to shall and to delete at a steady state power level of 1.0 MW is acceptable.

3.3 Proposed Addition of TS 5.3.1, Specification 4 As stated in its LAR, the licensee proposed to add Specification 4 to TS 5.3.1, which is denoted below using bold to indicate the proposed additional text.

4. G ring core locations may be empty (water-filled).

As stated in its LAR, the licensee proposed that G ring core locations may be empty, which would result in an empty (water-filled) position. As shown in figure 2 above, the G ring is the outer-most ring and contains 30 core positions. The licensee modeled several combinations of fuel elements, graphite elements, and empty (water-filled) positions to show that a transition of a single graphite element to an empty (water-filled) position in the G ring results in a power increase of 1 to 3 percent in the adjacent fuel element. The power increase is dependent on the adjacent fuel element type of either 20/20 or 30/20 and burnup level of the fuel element. Further, the licensee stated in its LAR that the transition of a second graphite element to an empty (water-filled) position in the G ring resulted in a change in power levels in adjacent fuel elements by +/- 1 to 2 percent. As stated in the LAR, the power levels in the G ring are on the order of 6 to 9 kW and each transition of a graphite element to an empty (water-filled) position reduces the excess reactivity by $0.05 units of reactivity.

NRC Staff Review The NRC staff reviewed the information in the LAR and finds that including an empty (water-filled) position in the G ring locations would increase the adjacent fuel element power level from a range of 6 to 9 kW to 6.1 to 9.3 kW for one empty (water-filled) position and 5.9 to 9.2 kW for two empty (water-filled) positions. The NRC staff notes that for close-packed fuel arrangements for TRIGA reactors, fuel elements in the outermost core ring produce lower power levels compared to the inner rings due to neutron losses at the core periphery. TS 5.3.1, Specification 2 requires that the fuel be arranged in a close-packed configuration (i.e., compact core with no unused or internal core positions that are water-filled) to ensure that excessive fuel element power will not occur. The NRC staff finds that the licensees analysis of the proposed core configuration to include empty (water-filled) positions in the G ring locations does not appreciably alter the power level of the adjacent elements and provides the licensee flexibility for configuring the core. Additionally, the NRC staff finds that by including one or two empty (water-filled) positions in the G ring, the power level of adjacent elements range from 5.9 to 9.2 kW, which is well below the 17.69 kW that is required by TS 4.1.4, Specification 3 to ensure that the previously approved T-H analysis in the SAR is bounding ensuring adequate cooling of the adjacent fuel elements.

The NRC staff finds that transitioning a graphite element to an empty position would reduce the overall excess reactivity that is used to compensate for fission poison build up, fuel temperature feedback, fuel burnup, and other core characteristics. In its LAR, the licensee states that 30/20 fuel elements would need to be placed in the inner-most core rings to ensure that sufficient excess reactivity exists. The NRC staff notes there is not a minimum TS limit for excess reactivity; however, enough excess reactivity is needed to compensate for certain core characteristics that result in negative reactivity feedback.

Based on the above, the NRC staff concludes that the licensees proposed new TS 5.3.1, Specification 4 allowing for G ring locations to be empty (water-filled) is acceptable.

3.4.

Conclusion Based on its review of the proposed changes to TS 4.1.4, Specification 3 and the proposed addition of TS 5.3.1, Specification 4, the NRC staff concludes that the TSs, as revised, continue to provide acceptable limitations on the maximum single fuel element power level that has been analyzed and on the core configuration of the G ring, respectively. These proposed TS changes would allow the licensee the operational flexibility to reconfigure the reactor core while maintaining safety. The NRC staff also concludes that the TSs, as revised, continue to comply with 10 CFR 50.36(c)(3) and 10 CFR 50.36(c)(4).

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement is required for any action within the categories of actions listed in 10 CFR 51.22(c). The Commission has declared these actions to be categorical exclusions by finding that the actions do not individually or cumulatively have a significant effect on the human environment.

4.1 Proposed Changes to TS 4.1.4, Specification 3 and Proposed Addition of TS 5.3.1, Specification 4 Section 51.22(c)(9) of 10 CFR, states, in part, that issuance of an amendment that changes a requirement with respect to the installation or use of facility components located within the restricted area, as defined in 10 CFR Part 20, meets the definition of a categorical exclusion, provided that the proposed changes satisfy each of the criteria listed below:

(i) The amendment or exemption involves no significant hazards consideration;

[10 CFR 51.22(c)(9)(i)]

Pursuant to 10 CFR 50.92, Issuance of amendment, paragraph (c), the Commission may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or [10 CFR 50.92(c)(1)]

As discussed in section 3 of this safety evaluation, the proposed changes to TS 4.1.4, Specification 3 would allow the licensee to reconfigure the core based on the analyzed single hottest fuel element power level. The proposed TS would continue to ensure that the maximum fuel element power level remains below 17.60 kW, which level was previously approved by the NRC. Therefore, the proposed TS would continue to ensure that the initial conditions in the accident analysis described in chapter 13 of the facility SAR remain valid. Adding Specification 4 to TS 5.3.1 allows for flexibility in core configuration by allowing water-filled positions in the G ring locations while contributing to a very slight increase in power level to adjacent fuel elements.

The proposed amendment would not change the analyses in SAR chapter 13, wherein the licensee previously analyzed a postulated maximum hypothetical accident that bounds all accidents at the facility, including ramp and rapid reactivity insertion events. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated, as stated in the licensees technical evaluation. The SAR analyzed reactivity accidents, loss of reactor coolant, and fission product release from clad rupture, and these accidents are not affected by the proposed change.

Further, no changes are being proposed to the reactor design or hardware, or to structures, systems, and components (SSCs) that are relied upon for accident detection, mitigation, or response. In addition, the proposed changes do not change the licensed power level of the reactor, the amount of special nuclear material or byproduct material authorized to be possessed and used at the facility, or any potential release paths from the facility. Therefore, the NRC staff finds that there is no significant increase in the probability or consequences of an accident previously evaluated.

(2) create the possibility of a new or different kind of accident from any accident previously evaluated; or [10 CFR 50.92(c)(2)]

The proposed changes to TS 4.1.4, Specification 3 and the addition of TS 5.3.1, Specification 4 do not create any new or different accident from any accident previously evaluated. The proposed changes do not involve any design or hardware changes to SSCs that are relied upon for accident detection, mitigation, or response.

In addition, the proposed changes would not introduce any new accident scenarios, transient precursors, failure mechanisms, or limiting single failures, and there would be no adverse effect or challenges to any reactor safety-related systems as a result of the proposed changes. Therefore, the NRC staff finds that the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) involve a significant reduction in a margin of safety. [10 CFR 50.92(c)(3)]

The proposed changes to TS 4.1.4, Specification 3 and the addition of TS 5.3.1, Specification 4 do not authorize any changes in the design, function, or operation of SSCs, or change the authorized steady-state reactor power level. The proposed changes do not alter how safety limits or limiting safety system settings are determined, change the limiting conditions for operations, or adversely affect the reliability of equipment assumed to mitigate accidents in the facility. In addition, the proposed changes would not adversely affect equipment required to safely shut down the reactor and to maintain the reactor in a safe shutdown condition.

Therefore, the NRC staff finds that the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, the NRC staff concludes that the proposed amendment authorizing changes to TS 4.1.4, Specification 3 and the addition of TS 5.3.1, Specification 4 involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and [10 CFR 51.22(c)(9)(ii)]

The proposed changes to TS 4.1.4, Specification 3 and the addition of TS 5.3.1, Specification 4 would not result in a significant change in the types or a significant increase in the amounts of fission products and effluents generated by operation of the reactor.

License Condition 2.C.1, Maximum Power Level, of Renewed Facility Operating License No. R-130 would continue to limit reactor operation to a steady-state maximum power level of 1,000 kW (thermal) power and the proposed amendment would not change potential release paths from the facility or the amount of nuclear materials authorized to be possessed and used at the facility. The operation of the building ventilation system, as required by TS 3.5, Ventilation and Confinement System, would continue to control the amounts of effluents that may be released off site and provides reasonable assurance that releases do not exceed the limits in 10 CFR Part 20, Table 2, Appendix B. Therefore, the NRC staff finds that there is no significant change in the types or significant increase in the amounts of any effluents that may be released off site.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

[10 CFR 51.22(c)(9)(iii)]

The proposed changes to TS 4.1.4, Specification 3 and the addition of TS 5.3.1, Specification 4 do not significantly affect individual or cumulative occupational radiation exposure. The amendment would not change the amount of nuclear material possessed and used at the reactor, the maximum power level, or postulated accident doses. Occupational and individual doses would remain below the limits in 10 CFR 20.1201, Occupational dose limits for adults, and 10 CFR 20.1301, Dose limits for individual members of the public.

Additionally, existing TS 6.4, Procedures, requires administrative controls for operating procedures that would continue to help ensure the adequacy of personnel radiation protection helping to limit individual or cumulative occupational radiation exposure.

Therefore, the NRC staff finds that there is no significant increase in individual or cumulative occupational radiation exposure.

4.2 Conclusion Based on its review, the NRC staff has determined that the issuance of the proposed amendment changes requirements with respect to the installation or use of facility components located within the restricted area under 10 CFR Part 50. The NRC staff has also determined that the proposed amendment involves no significant hazards consideration as well as no significant increase in the amounts, and no significant increase in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Therefore, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: G. Wertz, NRR M. Balazik, NRR Date: September 22, 2023