ML21092A137

Oregon State University - Regulatory Audit Re License Amendment Request to Remove Technical Specification Requirements

ML21092A137
Person / Time
Site:	Oregon State University
Issue date:	04/09/2021
From:	Michael Balazik NRC/NRR/DANU/UNPL
To:	Reese S Oregon State University
Balazik M
References
EPID L-2020-NFA-0005
Download: ML21092A137 (9)

Text

April 9, 2021 Dr. Steven R. Reese, Director Oregon State University 100 Radiation Center Corvallis, OR 97331-5903

SUBJECT:

OREGON STATE UNIVERSITY - REGULATORY AUDIT RE: LICENSE AMENDMENT REQUEST TO REMOVE TECHNICAL SPECIFICATION REQUIREMENTS RELATED TO THE INSTRUMENTED FUEL ELEMENT AND GRAMMATICAL CHANGES (EPID L-2020-NFA-0005)

Dear Dr. Reese:

By letter dated June 17, 2020 (Agencywide Documents Access and Management System Accession Nos. ML20171A576, and ML20171A575), Oregon State University (OSU) applied for an amendment to Facility Operating License No. R-106 for the Oregon State University TRIGA nuclear research reactor. The requested licensing action would amend the facility technical specifications (TSs) to remove the TS requirements related to the instrumented fuel element.

Additionally, OSU proposes to make numerous grammatical and editorial changes to the TS.

The U.S. Nuclear Regulatory Commission (NRC) staff will conduct a virtual regulatory audit starting on April 15, 2021, to gain a better understanding of the application and any proposed changes to the facility. The audit may include review of documentation, observation of the facility, and discussions with facility personnel and management. The enclosed audit plan provides additional details of the objective and scope of the audit. To facilitate an efficient audit, please provide ready access to working space, requested documentation, and areas of the facility.

Following completion of the audit, the NRC staff will provide an audit summary. The summary will include a description of any information identified during the audit that will need to be docketed to supplement the application and allow the NRC staff to continue its review.

S. Reese 2 If you have any questions, please contact me at (301) 415-2856, or by electronic mail at Michael.Balazik@nrc.gov.

Sincerely, Signed by Balazik, Michael on 04/09/21 Michael F. Balazik, 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. 50-243 License No. R-106

Enclosure:

As stated cc: See next page

Oregon State University Docket No. 50-243 cc:

Mayor of the City of Corvallis Corvallis, OR 97331 Maxwell Woods, Assistant Director Nuclear Safety and Emergency Preparedness Division 550 Capitol St. NE Salem, OR 97301 Dr. Irem Tumer, Vice President for Research Oregon State University A312 Kerr Administrative Services Bldg Corvallis, OR 97331-5904 Mr. Robert Schickler Reactor Administrator Oregon State University 100 Radiation Center, A-100 Corvallis, OR 97331-5903 Mr. Daniel Harlan, Chairman Reactor Operations Committee Oregon State University 100 Oak Creek Building Corvallis, OR 97331-5904 Test, Research and Training Reactor Newsletter Attention: Amber Johnson Dept of Materials Science and Engineering University of Maryland 4418 Stadium Drive College Park, MD 20742-2115

ML21092A137 NRR-106 OFFICE NRR/DANU/PM NRR/DANU/LA NRR/DANU/BC NRR/DANU/BC NAME MBalazik NParker DHardesty MBalazik DATE 4/6/2021 4/6/2021 4/7/2021 4/9/2021 OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY AUDIT PLAN REGARDING PROPOSED AMENDMENT NO. 26 TO RENEWED FACILITY OPERATING LICENSE NO. R-106 OREGON STATE UNIVERSITY OREGON STATE UNIVERSITY TRIGA NUCLEAR RESEARCH REACTOR DOCKET NO. 50-243

Background

The U.S. Nuclear Regulatory Commission (NRC) staff is continuing its review of the Oregon State University (OSU) TRIGA research reactor (OSTR) license amendment request (LAR),

provided by letter dated June 17, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20275A267). The LAR proposes to remove the technical specification (TS) requirements related to the instrumented fuel element (IFE). Additionally, OSU proposes to make numerous grammatical and editorial changes to the TSs. This regulatory audit is intended to assist the NRC staff in its review of the LAR.

Regulatory Audit Bases The purpose of this audit is to determine if the licensees proposed LAR meets all the applicable regulatory requirements of Title 10 of the Code of Federal Regulations and addresses applicable guidance provided in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Part 1, Format and Content, and Part 2, Standard Review Plan and Acceptance Criteria (ADAMS Accession Nos. ML042430055 and ML042430048, respectively).

Regulatory Audit Scope The NRC staff plans to conduct this virtual audit via teleconference and/or video conference with OSU staff. The NRC staff will review the OSUs LAR along with supporting basis and reference documentation. This audit will provide information necessary to continue the NRC staffs evaluation of the LAR. In addition, the audit may identify additional information that will be required to be docketed to support the basis of the licensing decision.

The NRC staff will use the online reference document portal provided by OSU staff. Access to the online portal is limited to specific NRC staff (e.g., based on NRC e-mail addresses or the use of passwords which will only be assigned to NRC staff directly involved in the LAR review on a need-to-know basis), and the documents in the online portal are read-only to prevent NRC staff from saving, copying, downloading, or printing any documents. The conditions associated with the online reference document portal must be maintained throughout the review process. The NRC staff who should be granted access to the portal are those listed in the Audit Team section below. The NRC staff will provide a request to close the online reference document portal at the conclusion of the audit.

Enclosure

2 Audit Information OSU should be prepared to support the NRC staff by having technical staff available for discussions and documentation supporting the LAR readily available. Additionally, the OSU staff should be prepared to provide referenced documents and reports, calculations, and computer code verification that support the analysis documented in the safety analysis report (SAR), bases for TSs, or rationale for any required plans and procedures, as necessary. The NRC staff will focus on the following information needs:

1. NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Standard Format and Content for Technical Specifications for Research Reactors, Section 4.1, Reactor Core Parameters, item (3), Pulse Limits, states the relationship between pulse peak fuel temperature and inserted reactivity should be determined when changes are made in the core.

The proposed TS reactivity limit of $2.30 provides a $0.03 margin to the recommended maximum fuel temperature limit of 830 degrees Celsius (1526 degrees Fahrenheit) during pulse mode operation. A minimal change in core reactivity may result in the fuel exceeding the maximum temperature limit as recommend by the fuel vendor. The OSTR TSs do not specify a surveillance requirement to determine the relationship between pulse peak fuel temperature and inserted reactivity when changes are made to the core.

Provide an explanation why a surveillance requirement to determine the relationship between pulse peak fuel temperature and inserted reactivity is not needed when changes are made to the core

2. During the conversion from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel, OSU stated, in part, in its response, dated June 20, 2008 (ADAMS Accession No. ML082350345), to the NRCs request for additional information (RAI), item 11, the following: The transient rod calibration (measured and simulated) consisted of a total of eight individual measurements. The MCNP5 error in total rod worth is ~5.5%. The measurement error in total rod worth, assuming eight measurements each with an uncertainty of 5% is ~14.1%.

Explain whether this measurement error of the transient rod was considered in the analysis of the proposed pulse limit of $2.30, when a pulse of $2.33 results in the maximum fuel temperature limit of 830 degrees Celsius (1526 degrees Fahrenheit).

3. The prompt temperature coefficient becomes less negative as the core ages, as shown in Figure 4-21 of the OSU HEU/LEU fuel conversion SAR (ADAMS Accession No. ML080420546). With a less negative prompt temperature coefficient, explain the resultant change in pulse power and maximum fuel temperature for the same reactivity insertion. Explain what other core characteristics that could change over core life that may affect the maximum fuel temperature for the same reactivity insertion for pulse mode operation. As these other core characteristics change over core lifetime, clarify whether a $2.30 pulse at core end-of-life would exceed the 830 degrees Celsius (1526 degrees Fahrenheit) temperature limit.

3

4. In the LAR, Pulse Mode Analysis, OSU states, in part, that although RELAP-3D (Reactor Excursion and Leak Analysis Program) is a deterministic code that does not provide uncertainty, the value of $2.30 was chosen as the pulse limit based upon the understood conservatism. However, the understood conservatism was not explained in the LAR. Describe and provide an explanation of the understood conservatism in the analysis for the determination of the $2.30 pulse limit.

5. For pulse mode operation, clarify any reactor parameters readily available to the operator to verify the inserted pulse reactivity does not exceed the proposed $2.30 limit thereby ensuring the maximum fuel temperature of 830 degrees Celsius (1526 degrees Fahrenheit) is not exceeded. For any reactor parameters identified, describe the associated instrumentation along with any surveillances.

6. TS 4.1, Reactor Core Parameters, specification e, states that 20 percent of the fuel elements comprising the core shall be inspected visually for damage or deterioration and measured for concentric or other swelling annual. Clarify whether the fuel elements immediately adjacent to the transient rod have were measured and visually inspected since the last pulse operation. If these fuel elements were inspected, provide the dates and results of the inspection.

7. Describe how the Limiting Core Configuration (LCC) was determined for this LAR.

Additionally, describe how future core alterations are confirmed to be within the analysis of record LCC.

8. Explain any differences between how the IFE is modeled in the reactor and thermal-hydraulic analyses compared to the standard TRIGA fuel elements in the core. If any differences are identified, explain how the events examined in OSUs SAR, Chapter 13, Accident Analysis, are impacted by replacing the IFE with a standard TRIGA fuel element.

9. With the proposed removal of the IFE requirements from TS, the protective functions based on Safety Power and Percent Power scrams are the primary limiting safety system settings (LSSS) safety functions. The NRC staff needs to understand all the details associate with these setpoints.

a. What is measurement uncertainty for the Safety Power and Percent Power systems and how were they determined?

b. What is the scram setpoint uncertainty for the Safety Power and Percent Power channels?

c. What operational uncertainty is assumed to account for normal core power fluctuations?

d. Describe the setpoint methodology used to set the LSSS scram functions and where the above uncertainties are included in that methodology.

4 Background Information

1. Provide the RELAP-3D thermal-hydraulic and reactor analyses available for review during the audit.

2. For the SAR Chapter 13 events:

a. Are there any scenarios that utilize operational assumptions rather than operation at TS limits? If so, explain why this is acceptable for bounding accident analysis calculations. Of the events analyzed, which event has the minimum margin to the TS limits?

b. Insertion of reactivity Describe how the mechanical block and electrical interlock for the transient rod is set to prevent the insertion of reactivity from exceeding $2.30 thereby limiting maximum fuel temperature below 830 degrees Celsius (1526 degrees Fahrenheit) as stated in TS 3.1.4, Pulse Mode Operation.

c. Loss-of-coolant accident (LOCA)

Describe the analytic geometry used the LOCA decay heat loss analysis.

3. Provide a comprehensive list of all the computer codes that were used in SAR Chapter 4, Reactor Description and Chapter 13 analyses and generally describe what those codes were used for, and whether those codes are new or modified from those used in the latest SAR.

Team Assignments The NRC staff performing this audit will be:

Michael Balazik (Audit Leader and Project Manager)

Jeremy Dean (Technical Reviewer)

Logistics The virtual audit will start on Thursday, April 15, 2021, with an entrance meeting, consisting of a focused review between the NRC staff and OSU staff. Following the review on April 15, 2021, the audit will continue as necessary until NRC staff have adequate understanding of issues to be addressed to complete the review of the LAR. The audit activities will be conducted via teleconference and/or video conferences, as appropriate, to support the efficient gathering of information by the NRC staff.

Audit Questions The questions for discussion during the audit are primarily based on open items listed above related to OSUs LAR. At the conclusion of the audit, necessary RAIs will be drafted to ensure the licensee has a clear understanding of the RAI scope.

5 Deliverables At the completion of the regulatory audit the NRC staff will prepare a regulatory audit summary, which will be issued within 90 days after the audit. The regulatory audit summary will include the documents reviewed, the audit activities, and any RAIs that were discussed or that will be issued based on the audit, as applicable.

Audit Schedule The NRC staff will conduct teleconferences and/or video conferences with OSU staff starting on April 15, 2021. Additional audit interaction will continue remotely as necessary.

Earlier if possible but no later than April 15, 2021, provide NRC staff access to the online reference documents portal with the documents stated in Audit Information above.