ML22333A623

University of Texas, Submittal of 14-Day Report for Event Number 56198

ML22333A623
Person / Time
Site:	University of Texas at Austin
Issue date:	11/16/2022
From:	Charlton W University of Texas at Austin
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
56198
Download: ML22333A623 (1)

Text

WALKER DEPARTMENT OF MECHANICAL ENGINEERING Nuclear Engineering Teaching Laboratory Pickle Research Campus R-9000

• Austin, Texas 78758

• 512-232-5380

• FAX 512-471-4589 n11clear.enf!:r. utexas. edu

• wcharlton(a)pustin. 11/e.xas. edu November 16, 2022 US Nuclear Regulat ory Commiss ion (NRC)

Washington, D.C. 20555 Attention : Document Control Desk

SUBJECT:

Docket 50-602, 14 Day Report Sir:

In accordance wit h t he Technical Specifications for The University of Texas Nuclear Research Reactor, enclosed is a 14-day report for event number 56198 reported to the Headquarters Operations Officer on November 2, 2022 .

Sincerely, t//'~

William S. Charlton , Ph.D .

Direct or, Nuclear Engi neering Teaching Laboratory

14-DAY REPORT: 2022 ALUMINUM FUEL EVENT AND FOLLOWUP William S. Charlton (Director), P. Michael Whaley (Associate Director), and James Terry (Acting Reactor Manager)

Nuclea r Engineering Teaching Laboratory (NETL)

University of Texas at Austin (UT-Austin)

Austin, TX 78758 November 16, 2022 1 INTRODUCTION On Octobe r 18, 2022, the University ofTexas at Austin (UT-Austin) notified the U.S. Nuclear Regulatory Commission (NRC) t hat the UT-Austin Nuclear Engineering Teaching Laboratory (NETL) staff had identifi ed potential non-compliances with Technical Specifications while shutdown and had suspended operati ons. On October 17, 2022, NETL operat ions staff discovered that two (2) aluminum-clad standard TRIGA fuel elements had been loaded into the core (in locations C6 and D1) on January 4, 2022 and th e reactor had operated with those elements in an unana lyzed condition from January 6, 2022 to October 17, 2022 . Th is is a non-compliance of the NETL technical specifications wh ich states 1:

Definition 1.5, Fue l Elements, Standard : "A fuel element is a single TRIGA element of standard type. Fuel is U-ZrH clad in sta inless steel clad . Hydrogen to zirconium ratio is nominally 1.6."

LCO 3.1.4, Fuel Elements part a. " In measuring the elongation, the length exceeds the original length by 2.54 mm (1/10 inch) ." and part b. " In measuring the transverse bend, the bend exceeds the original bend by 1.5875 mm (1/16 inch)." 2 Design Section 5.3.1 Fuel Elements part c.: "Cladding: 304 stainless steel, nominal .020 inches thic k."

Bases A.3 .1.4: "The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequat e cooling flow . The limit of transverse bend has been shown to result in no difficulty in disassembli ng the reactor core."

Upon discove ry, th e reactor was immediately shutdown, the fuel elements were removed, and the elements were inspected for damage. There was no damage to the elements. The fuel elements are safe ly stored at the facil ity.

Ana lysis has indicated that the fuel rema ined below the safe temperature limit for its design during all operat ions from January 6, 2022 to October 17, 2022 and that operation with this fuel could not have caused the exist ence or development of an unsafe condition with regard to reactor operations .

1

" Fa cil ity Operati ng Licen se, Docket No. 50-602, University ofTexas at Au st in, Licen se No. R-129," US Nuclear Regulatory Commission, Agencywide Documents Acce ss and Management System [ADAMS] Accession No. ML14136A073 (December 1990).

2 NOTE: NUREG-1537 specifies a larger tolerance, at 1/8 in . for both bend and length elongati on 1

The Limiting Safety System Settings {LSSS) for the NETL reactor are :

1. a maximum temperature monitored by instrumented element in the B or Cring of 550 °C,

2. a maximum steady-state power level of 1.1 MW, and

3. a maximum transient (pulsed) reactivity insertion of 2.2% Lik/k.

Analyses were performed for norma l operation (pulsing and non-pulsed) with the elements in the C6 and Dl location, normal operations with one of the elements in the hottest location of the core (the Bl posit ion), normal operation with the alum in um clad elements in the hottest location of the core (the Bl position) and the two instrumented fuel elements {IFE' s) in the coldest location allowed by technical specifications (positions C8 and C9), and under design basis accident conditions (Loss of Coolant Acci dent) . In all conditions the fuel was protected by the high power scram or by the facility design maintaining peak aluminum clad fue l temperature below 500°C.

2 EVENT TIMELINE In 2004, a sh ipment of fuel was received from the University of Illinois at Urbana-Champaign {UIUC) that included 2 aluminum clad fuel elements, which were placed in spent fuel storage. An informal tracking mecha nism, using an MS Excel file titled "B159.xls", was updated to reflect the elements received including notation that 2 elements are aluminum SFE (standard fue l elements). Note that at the time, NETL staff were awa re of the aluminum-clad fuel elements not being useable in the NETL core, but accept ed receipt of the elements, we believe, because that was necessary in order to acquire the useable stainless steel clad elements which were also included in the shipment. It is believed that the intention was to sh ip the alum inum clad fuel elements to Idaho for disposal at some later date.

In 2018, UT received 2 shipments {19 elements each) of lightly irradiated TRIGA fuel elements from spe nt fuel interim storage at INL.

Also in 2018, all irradiated fuel inventory that was in the wells was moved to the reactor pool racks (except for 2 canned elements stored in a separate well) in anticipation of pending shipment to DOE (this included the two aluminu m-clad standard fuel elements which were present in NETL inventory).

Representatives of t he I NL interim storage facility performed onsite inspection of irradiated fuel designated for return to DOE as spent fue l in August 2018. During the INL inspection to support spent fuel shipment, NETL operations staff performed all fuel handling during the inspection and the aluminum clad elements were identified and documented in the inspection report as such .

In January 2022, the bienn ia l fue l inspection requ ired by Techn ical Specification was performed (completed January 4, 2022) . Following the fuel inspection, 10 fuel elements, including the two aluminum fuel elements, were installed in the core on January 4, 2022 in a campaign to increase excess react ivity. These aluminum-clad elements were chosen for insertion because they were listed on the 8159 .xls file as havi ng a low burn up (in column O of the "Historic Fuel data" tab in that file) . The B159 .xls file also list s (in column B of the same tab) the notation " Al SFE" intending to specify that the elements we re alum inum-cla d elements. This notation was missed by the operations staff when choosing t hese low burnup elements to insert into the core . The NETL procedures do not include instructions to verify that elements being moved into the core or measu red are only stainless steel cladding. NETL procedures do not require an independent review by NETL management that the core 2

load ing is only w ith qualified fuel. Thus, a mistake by a single individual following the approved NETL procedures led to errors in the loading of the core, and there was not a procedure in place to identify that error prior to reactor startup in January 2022 .

The location of the aluminum-clad elements present in the core was discovered on October 17, 2022 by the Acting Reactor Manager {the previous Reactor Manager had res igned from UT-Austin as of September 9, 2022 ), and the situation was reported to the Associate Director. Reactor operations were suspended, pending resolution {identificat ion and completion of corrective action to restore the core to the requ ired configuration) . The Acting Reactor Manager removed the aluminum elements, performed visual examination of the elements, and verified current fuel inspection for all elements in the core was com pl eted on Janu ary 4, 2022 . However, a record of inspection was not immediately available for the elements inserted to increase excess react ivity; the previous reactor manager was contacted and confirmed inspect ion had been completed on January 6, 2022 as part of the fuel load process.

The discove ry of th e aluminum -clad fuel elements in the UT-NETL reactor occurred when the new Rea ctor Manager was reviewing and updating fuel records and procedures . The Reactor Manager was worki ng to update and improve upon documentation and procedures to be more consistent with his pl an for management of t he reactor facil ity. While updating the 8159 .xls file to reflect all element locations in the fue l movement log, he identified that the two aluminum -clad elements were in the react or core . Th is discovery was not procedure-driven but occurred as a result of the new Reactor Ma nager' s initiative t o update NETL records management. If not discovered in this way, the location of the aluminum-clad fuel elements in the NETL reactor core would not likely have occurred until annual maintenance was performed in January 2023 .

The investigation follow ing the identification of the non-compliance associated with operation using the aluminum-clad fuel identified procedural non-compliance issues . Following the 2018 fuel inspection of stand ard f uel elem ents made by electronic mea surements (st rain gage) as identified in the procedure, the electronic mea surements were discontinued based on successful testing of the go/no-go gauge

{requ ired by the procedure) to evaluate bend and an underwater camera and scale to measure elongat ion . The length and bend data we re not recorded as requ ired by the procedure (only if the element passed the bend and elongation test was recorded) . The procedure had not been revised to reflect the change in test method . Consequently a 50.59 evaluation was not implemented, although it wou ld likely have screened out the change for the need of prior NRC rev iew. The usage of the strain gauge probably wo ul d not have indicated that the two elements were alum inum clad instead of stainless steel, but t he failu re to cont rol procedure changes and compliance as required indicated a less tha n adequat e self-critica l attit ude w ith a failure in attention to detail. The Associate Director therefore directed a compre hens ive review of procedure and procedure performance to identify other potential issues.

3 REPORTING TIMELINE October 17, 2022 (~2:00 PM) : NETL Associate Director contacted via phone call the UT-NETL NRC Program Manager {Andrew Waugh and later Geoffrey Wertz) on the day of the discovery at approximately 2: 00 PM local ti me to inform them of the possible non-compliance and to discuss reporta bility of t he inci dent. Fol lowing that phone cal l and based on the available evidence and Tech nical Speci fi cations 6.6. 2, NETL determ ined that the incident wa s not reportable to the NRC, but UT-NETL staff would keep NRC informed as we proceeded through analysis, corrective actions, and 3

eventual restart. It was determined that the usage of aluminum-clad fuel in the C6 and D1 locations in the UT-NETL core could not have "caused the existence or development of an unsafe condition with regard to reactor operations" because the peak fuel temperature of the aluminum clad fuels would stay below 500°C in the event that the Instrumented Fuel Element {IFE) at B6 scrammed at 550°C and that at no time during pulsing operations would the peak fuel temperature anywhere in t he core exceed 420°C.

October 18, 2022 (8:34 AM): UT-NETL staff provided the initial version of a summary document to NRC Staff (Geoffrey Wertz and Andrew Waugh) . That summary detailed the incident, our initial root cause findings, our initial corrective action plan, and our review of reportability. NETL staff also informed the UT-Austin Reactor Oversight Committee (ROC) of the potential non-compliance issue and suspension of operations. The ROC was also provided a copy of the summary document.

October 21, 2022 (8:30 AM): UT-NETL staff met with US NRC staff (Geoffrey Wertz, Andrew Waugh, and Kevin Roche) to provide an update on the UT-NETL situation and continued analysis. This included updated MCNP and TRACE analysis of the fuel temperatures under steady-state and pulsing operations.

October 25, 2022 (3:36 PM) : UT-NETL staff provided an updated summary document to the UT-Austin ROC as well as a revised MAIN-5 procedure for review and approval.

October 26, 2022 (9:00 AM): UT-NETL staff provided an updated summary report of the event via email.

UT-NETL staff met with US NRC staff (Geoffrey Wertz, Andrew Waugh, Kevin Roche, and Travis Tate) t o provide an additional update on the UT-NETL situation, to step the NRC staff through the analysis in the most recent summary document provided, and answer questions regarding that summary update document.

October 26, 2022 (1:00 PM) : UT-NETL staff met with US NRC staff (Andrew Waugh, Kevin Roche, Travis Tate, Josh Borromeo, Mohamed Shams, and Jeremy Brown) to provide an additional update on the UT-NETL situation . NRC expressed concerns because the event involved an unanalyzed condition for t he UT-NETL core (note that this would be a reportable condition for a power reactor). This condition is not specified as reportable in the UT-NETL license, but the NRC expressed concerns that (1) the safety margin for the aluminum-clad fuel had not remain sufficiently large during operation and (2) that the inadequate controls at NETL "could have caused the existence or development of an unsafe condition with regard to reactor operations". NETL staff agreed to continue analysis of ossible ways in which an unsafe condition could have been caused. NETL staff agreed to seek guidance from both the UT Reactor Oversight Committee (ROC) and US NRC prior to commencing r estart operations to ensure the reactor is operated safely.

October 31, 2022 (1:00 PM) : UT-NETL staff provided updates to the ROC and met with ROC members to discuss the non-compliance issues and corrective action plan .

November 1, 2022 (11:30 AM) : UT-NETL staff met with members of the ROC to discuss the corrective action plan and reportability analysis including hypothetical "what if" scenarios.

November 2, 2022 (8:34 AM): UT-NETL staff contacted the NRC Headquarters Operations Center by t elephone and informed them of a potential reportable occurrence in which NETL had identified inadequacies in the procedures and administrative controls such that the inadequacies could have 4

caused the existence or development of an unsafe condition with regard to reactor operations.

After investigation of the event, development of the corrective action plan, and progress on implementing the corrective action plan, NETL staff identified changes in procedures and administrative controls that would have been adequate to prevent insertion of disqualified fuel in NETL reactor core. While analysis has shown that the insertion of aluminum clad fuel did not have t he potential to cause the existence or development of an unsafe condition, insertion of other fuel disqualified by surveillance activities could potentially have caused the existence or development of an unsafe condition with regard to reactor operations. Thus, NETL management determined on November 1, 2022 that this could potentially be classified as a reportable event under the category "An observed inadequacy 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 reactor operations".

4 NON-COMPLIANCES We identified five non -compliance issues :

1. Fuel that is not specified for use at the UT-NETL reactor was used in the core

2. Procedure violation, fuel inspection method was not conducted in accordance with (IAW) procedure

3. Procedure violation, records of fuel inspection required by the procedure were not generated

4. Procedure violation, fa ilure to revise procedures IAW administrative controls (including 50.59)

5. Failure to complete a 50.59 evaluation for usage of aluminum clad fuel at the UT-NETL reactor The fi rst of these was identified immediately from the discovery of the two aluminum-clad fuel elements in the UT-NETL core. The remaining four non-compliances were discovered during the conduct of the cause analysis and subsequent internal investigations for this event.

5 SAFETY SIGN IFICANCE

1. PRIMARY - NETL operated from January 6, 2022 to October 17, 2022 with the reactor core in an unanalyzed condition . While calculations indicate this specific configuration could not have caused the existence or development of an unsafe condition, NETL's controls were not adequate to prevent t he use of the unqualified fuel elements in the core .

2. SECONDARY - 1st Line Supervisor Failure to follow procedures:

Fuel inspect ion procedure as written was not performed Change co ntrol process (that invokes 50.59) was not followed 5

3. MITIGATION a) The go/no-go test is commonly used in TRIGA facilities and is adequate to meet the Technical Specifications requirements and is consistent with the Technical Specifications basis b) Aluminum clad fuel is commonly used at TRIGA reactors, and there has been no evidence of fuel degradation since installation in January c) Although the values were not recorded, the length measurement was performed and there has been no evidence of fuel degradation d) Initial analysis indicates aluminum fuel does not exceed the temperature safety limit in the C6 or D1 positions for operations since installation 6 ROOT CAUSE A root cause analysis was performed on October 17, 2022 immediately following discovery of the aluminum-clad fuel elements in the UT-NETL core. That root cause analysis involved a holistic approach studying the broader questions on what led to conditions in which the UT-NETL insertion of aluminum-clad fuel elements into the core was possible and what opportunities were missed to prevent this event from occurring. That root cause ana lysis has continued to be updated as more information was made available. We identified the following causes for this event:

A. Proximate Cause

1. Inappropriate element selected for installation B. Root cause

1. Procedural inadequacy: Lack of administrative, procedural, or engineering controls designed to keep elements not qualified for use out of the core C. Contributing causes

1. Lack of attention to detail:

• Failure to recognize the aluminum elements labeled in the B159.xls file while selecting elements for use

• Failure to question why the elements with a high uranium content (exceeding most of the fuel elements previously to restoring core excess reactivity) were not in use

2. Procedure inadequacies:

• This event occurred in the context of procedures that were adequate when personnel experience was high (the previous reactor manager had been employed at UT-Austin since before the NETL was built)

3. Inadequate administrative or engineered controls

• No administrative or engineered barriers were implemented that segregate or limited operator access to disqualified fuel

4. Inadequate safety conscious work environment:

• This event occurred in the context of procedure revisions improperly implemented 6

• This event occurred in the context of noncompliance with administrative and technica l procedures

5. Lack of management oversight:

• This event was a single point failure that could have been prevented with a second check on planned utilization of fuel

• Management oversight and audits did not identify degradation of the safety conscious work environment 7 CORRECTIVE ACTION PLAN

1. Remove alum in um fuel element s Comp leted : 10/17/2022

2. Rev ise the surveillance procedure for fue l element inspection :

a. Remove t he stra in gage measu rements from the procedure and

b. Provide an approved alternative for t he measu rements Completed : 11/01/2022

3. Perform the revised surveillance for the core configuration prior to startup Status: App rova l complete 11/01/2022, fuel inspection in progress

4. Rev iew ot her procedu res that satisfy Technical Specifications surve illances, to evaluate if other non-complian ces have been introduced in performance Completed : 10/28/2022

5. Conduct control rod worth ca librations Status: Sched ule pending comp letion of fuel inspect ion

6. Include in t he B159 .xls file

a. Dat e of la st f uel inspection

b. A 'qua lified' or 'disqualified ' flag t o indicate fuel elements not to be used in the core Completed: 10/18/2022

7. Review the event with staff, emphasizing the importance of procedural compliance, the change cont rol process fo r procedures, the application of license and Technical Specifications as adm inistrative controls, and the incorporation of this into NETL culture 7

Status : TBD, prior to restart

8. Revise the fue l handling procedure to require fuel not in a tested configuration (i.e., not installed at t he last control rod worth calibration) to be verified prior to installation:

a. Qual ified/d isqua lified for use

b. Inspection comp leted within prior 2 years

c. Core loading only with qualified fuel verified by NETL management prior to startup Status: In progress

9. Develop a method to designate fuel racks with visible indications that the contents are not allowed to be used in t he core Status: In progress 8 FUEL TEMPERATURE ANALYSIS The analysis described in this section details the potential temperature of aluminum-clad fuel elements at a va riety of locations in the NETL core (including the C6 and D1 locations in which the two elements were located), under two core configurations (including the 113-element core configuration while the aluminum clad elements were in the core ), and under various operating conditions (including steady-state and pulsing as well as loss of coolant accident) to determine the potentia l for damage to alum inum clad f uel located anywhere in the core and under a variety of safety-related conditions. Analysis of the fuel temperature is required because alum inum -clad fuel has a lower acceptable fuel temperature than stai nless steel clad f uel. Aluminum-clad TRIGA fuel has been used in TRIGA reactors since the beginning of the TRIGA program s but with a lower limit on fuel temperature compared to stain less steel clad TRIGA fuel. NUREG-1537 states " For aluminum-clad UZrH 1.o LEU 8 w/o TRIGA fuel, NRC has accepted that t he peak fuel t em perature should not exceed 500°C" .3 8.1 Analysis for Pulsing Operations An assessment of peak fuel temperatures that occurred during pulsing operations while the aluminum fuel was loaded in the core was made based on TRACE calculations (as described in Appendix A and using Fu el Specifications from Appendix B). Peak temperatures for (1) the whole f uel matrix and (2) the location of the thermocouple were identified for a series of pulsed reactivit y calculations . The ratio of the element peak and the thermocouple peak temperatures was calculated for each pulsed reactivity. A co rrelation between the ratio and the IFE measured temperature was identified (Figure 1). The peak fuel temperature in the core for each $3 .00 pulse during calendar year 2022 was evaluated using the measured tempera t ure (from the IFE) and the corre lation . The results are shown in Figure 2. Thus, the pea k fu el t emperature was below 420°C in all fuel elements for all pulses in 2022.

3

" Guid el ines for Preparing and Reviewing App lications for the Licensing of Non-Power Rea ctors: Format and Content," U.S. Nuclear Regulatory Commission, NUREG -1537 Part 1, Appendix 14.1.

8

Ratio of Peak Fuel Temperature to IFE Reading Versus Measured Fuel Temperature 1.19 1.18 z\!J o

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Figure 1: Ratio of Peak Fuel Temperature to Indicated Fuel Temperature versus Indicated Fuel Temperature Calculated Using TRACE Peak Fuel Temperature and IFE Temperature for An $3.00 Pulses at UT-NHL Since Jan 2022 440 420 *

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The IFEs are located in the B ring, where the maximum power is generated . The aluminum fuel elements were inserted in the C and D rings, which generate less power than any elements in Bring and resu lt in lower temperatures. Since the peak fuel element temperature of the IFE was below 420°(, the values in the aluminum elements were significantly less than 420°C for all pulses with no potential for exceeding limiting temperature for pulsing aluminum elements . Th is also demonstrates that aluminum clad fuel element s located anywhere in the NETL core would not have exceeded the 500°C limit during any pulsing operations .

8.2 Analysis for Non-Pulsing Operations 8.2. 1 Normal Operation Up to 1.1 MW with Alumin um-Clad Fue l Elements in C6 and D1 For non-pulse operation, the LSSS setting is a temperature which, if exceeded, causes a reactor scram to be initiated preventing the safety limit from being exceeded. The UT-NETL Fuel Temperature LSSS is 550°( as measured in an IFE located in the B or Cring. During the time period of interest, NETL was operating with two IFEs (one in B3 and one in B6) with the LSSS at 550°C. An MCNP simulation was performed with the 113-element core configuration on January 6, 2022 with the two aluminum clad fuel elements in C6 and D1 to calculate power peaking factors (PF) for each fuel element location. PF is the ratio of peak fissio n energy in the element of interest to the fission energy in the average rod . The posit ions of primary interest to this analysis are the Bl, B3, B6, C6, CB, C9, and D1 positions. Bl and B6 are the hottest positions in the core . B3 is the position of the second IFE. C6 and D1 were the positions of the aluminum cl ad fuel elements from January 6, 2022 to October 17, 2022. CB and C9 are the coldest fuel positions in the C-ring. The power peaking factors for these locations are given in Table 1.

TRACE simu lations were used to acqu ire fuel element temperatures for fuel elements with various total element power. The temperature of a stainless steel clad or aluminum clad fuel element at a specific fuel element powe r is given in Table 2. A plot of the data in Table 2 is shown in Figure 3. As can be seen, t here is a significant decrease in element temperature for aluminum clad fuel compared to stainl ess steel clad fuel due to the much higher thermal conductivity of the 1100 aluminum alloy compared to stainless steel type 304.

Table 1. Power Peaking Factors for Positions of Interest in NETL Core Position Power Peaking Factor (PF)

Bl 1.750 B3 1.594 B6 1.734 (6 1.496 C8 1.418 (9 1.353 D1 1.395 10

Table 2. Steady-stat e fuel element temperature for stainless steel and aluminum clad fuel at varying fuel element powers.

Fuel Element Temperature (°C}

Fuel Element Power (kW} Stainless-Steel Clad Aluminum Clad 7.96 258 .7 212 .6 9.73 303 .8 247 .6 13.27 351.1 273 .6 13.78 359.7 279 .3 15.93 396.4 303 .8 16.84 411 .8 314.1 17.70 426.3 323 .9 19.47 455.8 344.0 21.69 492 .3 368.9 24.34 535 .7 398.6 25 .18 549.2 407 .9 25 .24 550.2 408.6 25 .31 551.4 409.4 25.44 551.6 410 .8 25.62 554.4 412 .8 25.66 555 .2 413 .3 26 .55 571.1 423 .0 28 .13 601 .7 442.9 29 .11 617 .9 454 .0 29.18 619 .1 454.9 29 .26 620.4 455.8 29.41 620 .9 457.6 29.62 624.3 459 .9 29.67 625.1 460 .5 30.69 644 .2 472 .2 31.86 680.6 495 .9 33.0 700.2 509.3 11

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• Aluminum Clad Figure 3: Expect ed Fuel Element Temperature Versus Element Power for Stainless Steel Clad and Aluminum Clad Standard TRIGA Fuel.

The neutronic and thermal hydraulic data produced above was used to calculate expected aluminum-clad fuel element temperatures for a variety of core configurations. The goal was to assess the possibility of an aluminum -clad fuel element exceeding the 500°C temperature limit in non-pulsing operations . The NETL core is protected by a high power scram (at 1100 kW) and a high temperature scram (at 550°C on either of the installed IFEs). The first two analyses are performed up to the 1.1 MW power limit at which a high-power scram would initiate. The last four analyses were performed assuming the high-power scram did not function and the core was only protected by the high-tem perature sc ram . These last four are potential hypothetical "what if" scenarios including one scenario in which the core configuration was changed to an 84-element core.

8.2. 2 Operat ion at up to 1.1 MW with Aluminu m-Clad Elements in C6 and D1 From t he MCNP simulation, the ratio of fission energy produced in the aluminum clad fuel elements at C6 and 01 to the fission energy produced in the IFE at B6 (the hottest IFE) was 0.863 and 0.804, respectively. During normal operation (which is 950 kW for the NETL), the IFE's measure fuel temperatures below 410°C. The TRACE simulation calculates a fuel temperature of a stainless steel clad IFE at B6 as 411.8°C at 1100 kW which corresponds to an average power per element of 9.73 kW (for a 113-element core) and a power in the B6 element of 16.84 kW. This calculated IFE temperature at 1100 kW of 411.8oC agrees well with the observed IFE temperatures below 410°C at 950 kW. If the IFE was 12

measuring a temperature of 411 .8°C (expected temperature at the licensed power limit of 1100 kW),

the n the fuel temperature in the aluminum clad fuel elements at C6 and 01 would be expected to be 287.9°( and 276.6°( , respectively . Thus, the aluminum clad elements during normal operation had fuel tem pe ratures we ll below the 500°( limit.

8.2 .3 Normal Operation Up to 1.1 MW with Aluminum -Clad Fuel Element in Bl If one of the alum inum-cl ad elements was located in the Bl position (wh ich is the hottest position in the core and essentially equal in power to the 86 position), then the aluminum clad element would have had a temperature of 315 .9°( at full license power of 1100 kW. This demonstrates that aluminum clad fuel elements located anywhere in the NETL core would not have exceeded the 500°( limit during any non-pulsing operations below 1.1 MW.

8.2.4 Operation in Excess of 1.1 MW with Alu minum-Clad Elements in C6 and D1 Sim ulations show that for the IFE at 86 to reach a temperature of 550°C in the current 113-element core configuration, the reactor power would be 1649 kW and the hottest element in the core would have a single element power of 25 .2 kW . If the fuel temperature at 86 in this core configuration increased to 550°C to initiate a reactor scram from the LSSS fo r fuel temperature, the temperature of the aluminum-clad fuel elements at C6 and 0 1 would reach 370.2°C and 353.7°C, respectively. Thus, even in the event of rea ching t he LSSS for f uel temperature, the aluminum -clad elements in C6 and 01 remained well below the 500°( limit for alum inum clad TRIGA fuel.

8.2.5 Operation in Excess of 1.1 MW with Aluminum -Cl ad Element in Bl This next analysis (and the two follow ing) was to determine the maximum possible temperature of an aluminum clad f uel element if located anywhere in the core at the point where the IFE would cause a high temperature scram (at IFE temperatu re of 550°(). At a tota l reactor power of 1649 kW, the IFE at 86 would reach 550°C and initiate a high temperature scram . If one of the aluminum -clad elements was locat ed in the Bl position (again the hottest position in the co re and essentially equal in power to the B6 position), then the aluminum clad element would have had a temperature of 411.2°( at a reactor power of 1649 kW. This demonstrates that alum inum clad fuel elements located anywhere in the NETL core wou ld not have exceeded the 500°( limit during any non-pulsing operations up to the initiation of a high tem pe rature sc ram with the IFE located at B6.

8.2 .6 Operation in Excess of 1.1 MW with Aluminum -Clad Element in Bl and IFEs in C8 and C9 Analysis was also performed placing one of the aluminum-clad elements in the Bl position (again the hottest position in the core and essentia lly equal in power to the B6 position) and placing one IFE in the C8 posi t ion and the other IFE in the C9 position (the lowest power element positi ons in the C-ring) . This is th e most limiting condition possible with the 113-element core configuration . While NETL always operat es wit h the IFEs in the B-ring, the LSSS states " a maximum temperature monitored by instrumented element in the B or Cring of 550°(" and the NETL license states in 3.2.3 Reactor Safety Systems that 2 ope rable Fuel Temperature Channels are required for operation in all modes. In this configu ration, the IFE at C8 will reach 550°( (initiating a reactor scram) prior to the IFE at C9. So this analysis was to predict the highest possible temperature that an aluminum-clad fuel element would be subj ect to the case t hat the IF Es were placed in the worst possible position in a 113-element core. If the IFE at C8 had a temperature of 550°C, the n the aluminum -clad element at Bl would have a temperature of 481.4°( which is only 19°C below the 500°( limit . Wh ile this is stil l be low the fuel temperature limit 13

for an aluminum-clad standard TRIGA fuel element, it does show a significant decrease in safety margin in this very limiting condition .

8.2.7 Operation in Excess of 1.1 MW with Alu minum-Clad Ele ment in Bl and IFE sin C3 and C7 in an 84-Element Core Th is analysis was to predict the highest possible temperature that an aluminum-clad fuel element would be subject to in the case that the IFEs were placed in the worst possible position in the most limiting core configuration possible. The Limiting Core Configuration (LCC) for the NETL reactor is an 84-element core . This is the configuration that provides just under the license limit of $7.00 core excess reactivity.

In this configuration, the hottest element is located at BS (with a PF=l.691) and the lowest power elements in the C-ring are at C7 (PF=l.220) and C3 (PF=l.338) . In this configuration, the IFE at C3 will reach SS0°C (initiating a reactor scram) prior to the IFE at C7. If the IFE at C3 had a temperature of SS0°C, then the aluminum-clad element at BS would have a temperature of 496.3°C which is only 3.7°C below the S00°C limit and within the margin of error of the TRACE code . Also, it should be noted that variations in the core loading could alter the PF va lues in this 84-element core. Thus, in this Limiting Core Configuration, it wou ld not be possible to ensure that an aluminum-clad fuel element would remain below the S00°C temperature limit.

8.3 Analysis for Loss of Coolant Accident (LOCA)

Th e inal analysis completed was to consider the temperature of the aluminum-clad fuel elements under a loss of coolant accident condition. The NETL core was simulated assuming full-power operations for 30 days (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day and S days per week for 30 days) and with the 113 element core. The reactor was then shutdown at initiation of the LOCA and heat was produced from decay heat. Peak fuel tem peratures for the aluminum-clad elements were calculated assuming the following four cooling con itions: (1) with water cooling for only 1 second following initiation of reactor SCRAM, (2) with water cool ing for on ly 60 second following initiation of reactor SCRAM, (3) with water cooling for only 600 seconds following initiation of reactor SCRAM, and (4) with water cooling for only 1200 seconds following initiation of reactor SCRAM . (Note that the NETL reactor pool contains over 11,000 gallons of water .) The calculated peak fuel tempe ratures versus time after SCRAM are shown in Fig. 4. Under the worst possible cool ing cond itions, the peak fuel temperature is just below 4S0°C and remains below 500°( for all cases considered .

14

Aluminum Fuel LOCA Response 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s/day, Sdays a week, 30 days, 113 Elements, 14.54 kW 500 450 u

~

400 -

~ 350

....::J f! 300

8. 250 E

{!!. 200 E

s 150 E

• xIQ 100

~ 50 0

0 2000 4000 6000 8000 10000 12000 14000 16000 Time After Air Cooling Initiated (s}

--H2O cool 1 s -H2O cool 60 s - H2O cool 600 s - H2O cool 1200 s Figure 4: Expect ed Fuel Element Temperature Versus Element Power for Stainless Steel Clad and Aluminum Clad Standard TRIGA Fuel.

8.4 Assessment from Analysis Cases MCNP and TRACE simulations were performed to assess the possible safety significance of operation of the NETL co re with two aluminum-clad fuel elements inadvertently inserted into the core. In the actual event that occurred, two partially burned aluminum-clad elements were inserted into the NETL core in posit ions C6 and D1 with the reactor in a 113-element configuration and with IFEs located at B3 and B6.

Analysis was performed for this configuration in pulsing and steady-state operation, for more limiting configurations based on potential hypothetical "what if" scenarios, and for a loss of coolant accident .

The fo llowing conclusions can be drawn from the results:

1. In pulsing operations with the current 113-element core configuration, the insertion of two aluminum-clad fuel elements in the hottest location of the NETL core would lead to a fuel element temperature below 420°C and well below the 500°C limit.

2. In steady-state operations with the current 113-element core configuration, the insertion of two aluminum-clad fuel elements in the hottest location of the NETL core and the IFEs located in the B3 and B6 positions wou ld lead to a fuel element temperature below 412°C and well below the 500°C limit.

3. In steady-state operations with the current 113-element core configuration, the insertion of two aluminum-clad fuel elements in the hottest location of the NETL core and the IFEs located in the C8 and C6 positions (the lowest power positions in the C-ring) would lead to a fuel element t emperature below 482°C and below the 500°C limit.

4. For loss of coolant accident with the cu rrent 113-element core configuration, the peak fuel temperatures for the two aluminum-clad fuel elements reached a maximum of under 450°C 15

with only 1 second of water cooling available after initiation of the reactor SCRAM. Thus the rods remained well below the 500°( limit.

5. In steady-state operations in the Limiting Core Configuration (LCC) of 84-elements, the insertion of two aluminum-clad fuel elem ents in the hottest location of the NETL core and the IF Es located in the C3 and C7 positions (the lowest power positions in the C-ring) would lead to a fuel element temperature below 493°C which is only slightly below the 500°C limit (and within calculational uncertainty of the methods used here).

All of the cases involving the NETL core configuration present on the dates of the event in question resu lted in aluminum-clad fuel element temperatures well below the 500°( limit. Thus, the event in question did not causes nor could it have caused the existence or development of an unsaf e conditi on with regard to react or operations.

Whi le all of the cases considered are below the fuel temperature limit for an aluminum-clad standard TRIGA fuel element, the cases in which the IFEs are placed in the C-ring (in the 84-element core) showed fuel temperatures could reach near 500°( and possibly over 500°C. Thus, it was concluded that insertion of aluminum-clad fuel in other core configurations would potentially have caused the existence or development of an unsafe condition with regard to reactor operations but not with the 113-element core present during the time of the event.

9 REVIEW FOR REPORTABILITY Techn ical Specifications 6.6.2, Special Reports, Criteria and Evaluation :

a. Operation with actual safety-system settings for required systems less conservative than the limiting safety system settings specified in the technical specifications.

The actual safety-system settings and limiting safety system settings specified in the technical specifications were as foll ows :

Setting Limiting safety system Actual safety-system settings specified in the settings technical specifications maximum temperature monitored by 550 °( 550 °(

instrumented element in the B or C ring maximum steady-state power level 1.1 MW 1.lMW maximum transient (pulsed) reactivity 2.2% t.k/k 2.1% t.k/k insertion Thus, Safety-System Settings were not affected . - Does not apply 16

b. Operation in violation of limiting conditions for operation established in technical specifications unless prompt remedial action is taken.

Fuel type is not a Limiting Condition for Operation; fuel type is defined and listed as a design specification while the Limiting Condition for Operation applies to fuel damage, and prompt remedial action was taken on discovery - Does not apply

c. A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns.

No Safety System malfunction, although discovered while shutdown - Does not apply

d. An unanticipated or uncontrolled change in reactivity greater than one dollar. Reactor trips resulting from a known cause are excluded.

No unanticipated or uncontrolled change in reactivity - Does not apply

e. Abnormal and significant degradation in reactor fuel, or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.

No degradation in fuel or cladding - Does not apply

f. An observed inadequacy 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 reactor operations.

The inadequacy in the implementation of procedural controls applies only to (1) the use of TRIGA aluminum-clad fuel and (2) a change in the method for testing fuel degradation . Analysis in section 8 showed that at no time could these inadequacies have caused the existence or development of an unsafe condition with regard to reactor operations.

This event (the insertion of aluminum clad fuel in the C6 and Dl positions) did not have the potential to cause the development of an unsafe condition w ith respect to protecting the health and safety of the publ ic or facility staff. We did not identify any inadequacies in implementation of administrative or procedural controls that could have caused the existence or development of an unsafe condition with regard to reactor operations. Therefore, this was not reportable as An observed inadequacy 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 reactor operations. However, while the implementation of the existing procedures and administrative controls was adequate, NETL staff identified on November 1, 2022 (in the course of the investigation into this event) inadequacies in the procedures and administrative controls themselves which do not adequately prevent 17

insertion of disqualified fuel in the NETL reactor core . For this reason, NETL management had chosen to consider this a reportable event under this category and had reported it to the NRC Headquarters Operations Center on November 2, 2022 .

The Technical Specifications Basis for acceptance of measurements on fuel (A.3.1.4 Fuel Elements) is:

The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow. The limit of transverse bend has been shown to result in no difficulty in disassembling the reactor core.

As a result of examining procedural compliance 10 elements were found to be installed in the core with no discernable reference for determining elongation in the working records, and no values recorded to support future elongation measurements. (Bend measurements use a go/no-gauge and were likely conducted as required) . Historical measurements show a clear trend of elongation for some elements progress with burnup. If the surveillance proceeded into the future as conducted (from 2018 to 2022) then internal stresses could increase undetected terminating in a possible loss of containment integrity. Therefore a notification was made of "An observed inadequacy 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 reactor operations" when the installation was identified and verified . Two of the elements with no prior measurements and no record of measurement supporting installation were aluminum clad . The direct cause of installation of the aluminum elements and the failure to conduct surveillance activity required to assure fuel element cladding integrity was inattention to detail and failure to follow the approved procedure . An unauthorized procedure was used that did not provide instruction adequate to meet the requirements of the Technical Specification surveillance, so a contributing cause was an administrative failure to control procedure revision in accordance with the approved process.

NETL staff was able to recover reference data from 2004 sh ipping records and other fuel measurements from the original decommissioned UT-Austin reactor. The working documents have been revised to reflect the reference measurement. All elements in the core are being exami ned and measured using a standard method and elements that will replace those disqualified for use will be measured (either to establish a reference value or to compare to a reference value, as appropriate) before installation.

10 CONCLUSIONS The insertion of aluminum-clad fuel elements in C6 and D1 could not have caused the existence or development of an unsafe condition with regard to either steady-state or pulsing reactor operations .

However, NETL administrative and procedural controls to mitigate the possibility of insertion of disq ua lified fue l into the UT-NETL core were found to be inadequate .

18

APPENDIX A- METHODS USED IN FUEL TEMPERATURE ANALYSIS For this analysis, the MCNP 6.2 code was used for calculating fission rates in individual fuel rods throughout the NETL core. This code has been used for a variety of simulations for the NETL core in the past including calculation of control rod worths, power peaking factors, neutron flux spectra in experimental locations, and gamma dose rates at experimental locations . The results from these simulations have been benchmarked to a variety of measured values in the past 5 years and most recently were used to calculate neutron flux spectra and axial variation in the UT NETL central thimble with comparison to measured values. Generally, the results from these simulations agree to within +/-8%

of the measured va lues. The MCNP in put decks used in this analysis were based on the input decks from these previous ana lyses with modifications for insertion of the aluminum clad fuel elements into the simulation. 4 Thus, we expect MCNP simulations for the aluminum-clad fuel event to similarly have an accuracy of within +/-8%.

Thermal hydraulic modeling of the UT TRIGA was performed with TRAC/RE LAP Advanced Computational Engine {TRACE) . Thermal hydraulic characteristics were developed from classical methods and corrections for UT TRIGA geometry using the computational fluid dynamics code FLUENT. Distribution of fi ssion activity was developed from transport calculations in MCNP 6.2 . TRACE is designed to perform best-estimate analyses of operational transients and accident scenarios by modeling physical geometry and t hermodynamic conditions. TRACE is the NRC's flagship thermal-hydraulics analysis tool conso lidating and extending the capabilities of NRC's 3 legacy safety codes: TRAC-P, TRAC-Band RELAP.

NETL staff have used TRACE over the past five-years for thermal-hydraulic analysis of the NETL core including in support of relicensing. TRACE results have generally compared favorably with measured values for the co re .5 Th e low channe l unit cell cross section is based on the typical fuel element geometry, as illustrated in Fig. Al (unit cell and the surrounding f uel elements) . Some unit cell locations in the grid plate have different structures . The central thimble is not fueled, the transient rod does not contain fuel, and the fuel followers (which are generally not fully inserted in the core) have 80% of the fuel mass contained in standard fuel elements . This analysis uses a hot channel and assumes no interaction between adjacent unit cells. Any interaction between unit cells with fuel and adjacent unit cells with less or no fuel contributes a la rger area where convection flow is the result of heat transfer from the fully fueled cell, resu lting in enhanced heat removal from the fully fueled cell. Thus, from this standpoint the analysis here is conservative. As illustrated, the unit cell analysis is based on a fuel element and the surrounding flow area (end fittings have more complex geomet ry) circumscribed by a hexagon with an inner radius of 1/2 of the pin-to-pin pitch. The complex geometries of the fuel element end fittings are approximated as hydrodynamic characteristics.

4 "Analysis of the Neutronic Behavior of the Nuclear Engineering Teaching Laboratory Reactor at the University of Texa s," Radiation Center report, Oregon State University (March 2021) .

5 P.M . Whaley and W .S. Charlton, "Thermal Hydraulics Analysis of the University of Texas (UT) TRIGA Reactor,"

Nuclear Engineering Teaching Laboratory report, University of Texas at Austin (October 17, 2022).

19

Zr FfllROD Figure Al: Flow Channel for UT TRIGA Fuel Elements Figure A2: TRACE Model TRACE analysis is based on modeling a set of representative TRACE components with characteristics specifi ed by the user to model the system. The UT TRIGA model uses Break, Pipe, Heat Structure, and Power components. These TRACE components were assembled as shown in Figure A2 to model the thermal hydraulic perfo rmance of the unit cell flow channel.

An exam ple of a comparison of TRACE steady-state calculation results to measured values is given in Table Al. This shows the fuel temperatures measured in the IFE's located at B03 and B06 in December 2015 compared to TRACE simu lations of the same core configuration. Comparison of the calculated and observed data ind icates TRACE can predict steady-state behavior with reasonable accuracy (generally within +/-8% of measured temperature values) .

20

Table Al : Trace Calculated and IFE Measured Fuel Temperature Comparison IFE TRACE  % DIFFERENCE FUEL ELEMENT INDICATED CALCULATED BETWEEN POSITION ELEMENT POWER (kW) FUEL TEMP FUEL TEMP MEASURED AND (oC} (oC} CALCULATED B03 10878 13.24 325 345 -6.15%

B06 10708 13.61 364 354 2.74%

Anot her example is shown in Figures A3 and A4 for TRACE transient results . The results of TRACE simulations for the NETL core for four pulse reactivity insertions are compared to measured data from over 300 historical pu lses conducted at NETL. While there is significant scatter in power level and temperature data with some outliers, the results overall show excellent agreement. Thus, we expect the TRACE simulations to provide a reasonably accurate estimate for the aluminum -clad fuel event .

2.SOE+07 ff,l 2.00E+07 jJl QFi' i l.S0E*O7

i" .P

2. Jl

'l!'

t ,~

0 e

~ l .OOE+07 I l"'

5.00E+06 O.OOE*OO 1 -"

$1.00 $1.50 $2.00 $2.50 S3.00 Pulsed Reactivity Addition ($)

0 Historical Observed Data C TRACE Figure A3: Peak Element Power Level versus Pulse Reactivity Addition from UT TRACE Calculation Compared to Observed Historical Data 21

450 400 350 E3oo I!!

e2so 0) a.

E 200

~

Jl 150 0..

100 D so 0

$1.00 $15() $2.00 $2.50 $3,00 Pulse Reactivity Addition($)

o Historical Observed Data o TRACE Figu re A4 : Peak Fuel Temperature versus Pulse Reactivity Addition from UT TRACE Calculation Compared to Historical Data 22

APPENDIX B - FUEL ELEMENT SPECIFICATIONS The design of standard stainless steel clad fue l utilized in the NETL is shown in Figure AS . Stainless steel clad elements used at NETL all have fuel al loy length of 38.1 cm . The characteristics of standard fuel elements are shown in Table A2 . Table 2 also lists nomina l characteristics for aluminum-clad TRIGA fuel.

As can be seen there are several notable differences including slightly lower uranium content, lower H content, use of a burnable poison disk, no Zr rod, shorter fuel length, and slight change in fuel and rod dimensions. However, the most critical difference for this analysis is the change in cladding material and cladding thickness. The aluminum clad fu el elements have a significantly thicker clad than the stainless steel clad fuel elements. This change in clad thickness and material could have a significant impact on heat transfer out of the fuel element. 1100 Alloy Aluminum has a lower melting point than stainless steel t ype 304, but 1100 Alloy Aluminum has a much higher thermal conductivity (222 W/m-K versus 16.2 W/m -K at 37°(). 6 For the analyses performed here, the fuel elements were modeled using the data in Table 2 with the exception that the fuel isotopics were changed to match the listed uranium content and burnup for each fuel element (including the low burn up given for the aluminum clad fuel) .

ST..AJNLESS STEEL TOP END OF ATTI~ El

~ Ml bl STAJNlESS 1 STEEL i\./BE:

CLADDING I THICKNESS 0.02 1N.

ZIRCONIUM HYORI.DE-8,5Wf%

URAN!

15 IN.

1.475 JN.

MOLYDiSC,.

0.08MM THI STAJNLESSSTEEL GRAPH!'fE O 3.~ IN, BOTTOM END FITTING Figure AS : TRIGA stainless-steel clad fuel element design used in analysis .

6 Therma l conductivity data fo r illustrative purposes only and was from MatWeb Material Property Data at http://www.matweb.com.

23

Table A2 . Nominal Characteristics of Stainless Steel and Aluminum Clad Fuel Elements Fuel Element Type 8.5% Stainless Steel Clad 8% Aluminum Clad Fuel *- moderator material U-ZrH,. U-ZrH,o Uranium content 8.5 wt% 8.0 wt%

mu enrichment 19 .75% 19.75%

Average mu per element 39 g 36 g Burnable poison None 1% samarium/ 99% aluminum wafers Poison wafer thickness N/A 0.13 cm Shape Cylindrical Cylindrical Length of fuel meat 38.1 cm 35.6 cm Diameter of fuel meat 3.63 cm 3.58 cm Outer diameter of element 3.75 cm 3.76 cm Zirconium core diameter 0.635 cm None Cladding material Stainless Steel 304 Type 1100 Al Cladding thickness 0.0508 cm 0.076 cm Graphite Slug Outer Diameter 3.63 cm 3.60 cm Upper Graphite Slug Length 6.60 cm 9.40 cm Lower Graphite Slug Length 9.40 cm 9.40cm Molybdenum disc thickness 0.08 cm N/A 24