ML21351A317

UC Davis MNRC Response to NRC Staff Request for Additional Information Regarding Licensing Renewal Application Letter Issued November 30th, 2021

ML21351A317
Person / Time
Site:	University of California-Davis
Issue date:	12/17/2021
From:	Frey W McClellan Nuclear Research Center
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML21351A317 (4)

Text

UNIVE RSITY OF CALIFORNIA, DAVIS BERKELEY

• DA VIS * !RVlNE

• LOS ANGELES

• MERCED

• RIVERSIDE

• SAN DIEGO

• SAN FRANCISCO {\  ;\ __________

SANTA BARBARA

• SANTA CRUZ McClellan Nuclear Research Center UNIVERSITY OF CALIFORNIA Davis (9]6)-614-6200 December 17th, 2021 US Nuclear Regulatory Commission 11555 Rockville Pike MS 12-D3 Ro.ckville, MD 20852-2738

SUBJECT:

UC Davis MNRC Response to NRC Staff Request for Additional Informa tion Regarding Licensing Renewal Application Letter Issued Novemb er 301\ 2021.

Enclosed are the UC Davis MNRC responses to the 3 requests for additional information issued by the NRC staff on November 30th 2021.

I have reviewed this submission and found it to be truthful and accurate. If there are any questions or concerns, please contact me directly.

I declare under penalty of perjury that the foregoing is true and correct executed on Decemb er 17th , 2021 .

Sincerely,

[ti¥ 0-"v Wesley Ft'fy PhD Facility Director

NRC Request for Additional Information Item #1}

The regulation, 10 CFR 50.9 requires information to be complete and accurate in all material aspects. The guidance in NUREG-1537, Part 1, Section 1.4, "Shared Facilities and Equipment, 11 states, in part, that the application should describe systems and equipmen t that are shared with facilities not covered by the SAR.

Further, NUREG-1537 states that any safety implications that result from sharing facilities or systems should be evaluated in and referenced to the appropriate chapter of the SAR.

The NRC staff review of the licensee's SAR, Section 1.4.3, "Shared utilities, 11 could not ascertain if any of the utilities (electric, natural gas, water, phone, internet, etc.) used to support the operation of the reactor facility were shared with other facilities not described in the SAR, and/or if any safety implications could result from any shared utilities.

Describe whether any utilities are shared with other facilities not covered by the SAR, and if any safety implications could result from the loss of any shared utilities, or justify why no additional information is needed.

UCD/MNR C Response: All utilities provided to the MNRC are done by external providers.

In contrast to many other research reactors the MNRC reactor is located in a standalone building. The only shared utility is electrical service that is shared with the building directly south of the MNRC which is owned by the McClellan Business Park. While offsite utilities are required to operate the facility, none are required to place the reactor in a safe and secure condition. The loss of offsite utilities cannot generate any accident scenario .

The two utilities of most interest to all reactors are electricity and water. In the event of loss of electrical power, the facility maintains a backup battery supply which provides power to the console, reactor magnets, and all reactor indication for 15 minutes. This gives the reactor operator ample time to shut the reactor down in a controlled manner and verity the reactor has been placed in a secure condition. No electrical service (onsite or offsite) is required to maintain this secure condition.

Offsite water service is not directly required to operate the reactor or during a LOCA. This is due to the very small loss of primary water from evaporation during 1 MW operations. Typically, it would take approximately 2 weeks to reduce the level of water in the reactor tank to below what is required by the technical specifications. In chapter 13 of the SAR the analysis shows that the reactor does not need additional water (presumably from an offsite source) to prevent unacceptable heating of the core during a complete LOCA event.

Loss of offsite water can occur and does indirectly effect reactor operations as offsite water provides the water for the evaporative cooling system (secondary system) that cools the primary cooling system of the reactor. In the event of the loss of offsite water to feed the secondary system, the reactor operator would have several hours to ascertaining the situation and place the reactor in a secure configuration.

In this scenario, before the primary coolant exceeded the technical specification limit, the reactor operator would receive numerous warnings and alarms (e.g. cooling tower low float, secondary low flow alarm, and primary tank temperatu re warning) over the course of several hours, which would provide ample time to shut the reactor down.

NRC Request for Additional Information Item #2)

The regulation, 10 CFR 50.9 requires information to be complete and accurate in all material aspects. The guidance in NUREG-1537, Part 1, Section 1.8, "Facility Modifications and History," states, in part, that the SAR should indicate if the facility has not undergone significant or safety-relat ed physical or operational modifications since it was initially licensed, or since the last renewal was issued and should reflect any significant modifications made to the non-power reactor.

The NRC staff review of the licensee's SAR, Section 3.1.3, "Protection by Multiple Fission-Product Barriers (Criteria 10-19}," "Criterion 13: Instrumentation and Control," finds that the licensee states that "Note that square wave and pulse mode are no longer utilized at MNRC. 11 However, facility modifications needed to be implemente d to prevent an inadvertent square wave or pulse operation were not able to be identified in the SAR by the NRC staff Provide the following information to support the NRC staff review:

a. Provide a description of the facility modifications planned or implemente d to prevent an inadvertent square or pulse operation of the MNCR.

b. Provide an assessment of the potential for an inadvertent square wave or pulse event, including any changes or updates needed to the descriptions of any accident scenarios provided in the LRA SAR, Chapter 13, "Accident Analysis. 11

c. Provide any changes to the proposed technical specifications, if applicable to the proposed facility modifications.

UCD/MNRC Response: The compressed gas (nitrogen) supply has been disconnected and locked out at the base of the transient rod drive. The compressed gas cyl inder itself has been disconnected and removed from the facility. These modifications make rapid ejection of the transient rod physically impossible.

Therefore, pulses and square waves cannot be physically achieved by any means.

The square wave and pulse mode buttons on the console have been disable so.that they cannot be physically depressed. Prior to this when either of these buttons was depressed a circuit was completed so that if the reactor was below lkW, the NM-1000 would be disabled and the NPP-1000 would be readied for the reactor to perform a pulse or square wave. If the reactor was above 1 kW, the 1 kW interlock would prevent the bypass of the NM-1000 and would generate a warning to the reactor operator that the reactor cannot enter into pulse or square wave mode.

By physically preventing the pulse and square wave buttons from engaging the reacto r cannot inadve rtently be placed in a mode where the NM-1000 is bypassed. Therefore, the 1 kW interlock surveillance will no longer be required. Current and future ope rators will be trained on the specifics of these modificatio n. All other references to pulse or square wave model will be removed from the technical specifications.

NRC Request for Additional Information Item #3)

The regulation, 10 CFR 50.9 requires information to be complete and accurate in all material aspects. The guidance in NUREG-1537, Part 1, Section 4.2.2, "Control Rods, states, in part, that the applicant should 11 discuss the design, composition, and reactivity effects of control rod followers . Section 4.1, "Introduction, 11 and Section 4.2.3.1, "Control Function, of the SAR indicate that fuel-follower sections consist of 20 or 30 11

weigh t percen t uranium. However, Section 7.3.1, "Control Rods, 11 of the SAR states that the weigh t percen t of uranium in the fuel is 20.

Section 4.6.4.3, "Operating Core Configuration (OCC), of the 11 SAR indicates that the operating core configuration {OCC) only contains fuel-followers with 20 weigh t percen t uranium. Changes made to the OCC to develop the limiting core configuration (LCC} are described in Section 4.6.4.4, "Future Cores and the Limiting Core Configuration (LCC}, of the SAR, but changes to 11 the composition of control rod followers are not specified. Analysis of the OCC and LCC provides the basis for hot-channel power used in steady -state therm al hydraulic analysis of the MNRC core, as well as contro l rod worth values and power peaking factors used in the accident analyses. The 30/20 fuel-followers contai n more uranium by weigh t than 20/20 fuel-followers and are expec ted to have a higher reactivity worth.

If control rod followers are loaded in the core that differ from those model ed in the SAR, then control rod worth in the core may excee d that used in the accident analysis, and power density and peaking factors achiev ed in the LCC may not be limiting. The NRC staff reque sts the licensee to:

a. Clarify what control rod followers are model ed in analysis of the OCC and LCC described in the SAR.

b. Describe the allowable loading of control rod followers in the core.

c. If the allowable loading differs from that model ed in analys is of the OCC and LCC, please provide additional basis to confirm that design and accident analysis remains applic able.

MNRC /UCD Response: In both the OCC and LCC only 20/20 fuel follow ed controls were analyzed for in terms of therm al outpu t and peaking factors. A 30/20 fuel follow ed contro l rod would undoubtable have a notab ly higher reactivity worth , which would result in a larger therm al outpu t and overall eleme nt peaking factor. Based on operational and modeling experience with standard 30/20 fuel elements along with the position of the fuel followed control rods, it is unlikely a 30/20 fuel follow ed control rod would produce a peaking factor beyond what is already analyzed for in chapte r 4 of the SAR. However, this configuration has not been specifically analyzed for.

At this time the MNRC is willing to comm it to a technical specification that restricts the use of fuel follow ed control rod to only the 20/20 type. If the M NRC wishes to utilize a differe nt type of fuel follow ed control rod in the future , a license amen dmen t will be submi tted to the NRC for approval. MNRC has no intensions of reducing the total numb er of control rods in the core or their location.