ML17038A272
| ML17038A272 | |
| Person / Time | |
|---|---|
| Site: | Kansas State University |
| Issue date: | 03/28/2017 |
| From: | Traiforos S NRC/NRR/DPR/PRLB |
| To: | Geuther J Kansas State University |
| Traiforos S, NRR/DPR/PRLB, 301-415-3965 | |
| References | |
| CAC A11010 | |
| Download: ML17038A272 (8) | |
Text
March 28, 2017 Dr. Jeffrey A. Geuther, Manager Kansas State University Nuclear Reactor Facility 112 Ward Hall Manhattan, KS 66506-5204
SUBJECT:
KANSAS STATE UNIVERSITY - REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR FACILITY OPERATING LICENSE NO. R-88, FOR THE KANSAS STATE UNIVERSITY NUCLEAR REACTOR (CAC NO. A11010)
Dear Dr. Geuther:
The U.S. Nuclear Regulatory Commission staff is continuing its review of your license amendment application dated April 9, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12109A063), as supplemented by letters dated April 28, 2014, and October 5, 2016 (ADAMS Accession Nos. ML16200A317 and ML16291A498, respectively), to Facility Operating License No. R-88 for the Kansas State University (KSU) Research Reactor, to allow the use of up to four fuel elements of 12.5 percent uranium by weight in certain locations of the core of the KSU Research Reactor.
During our review of your responses and further reviews of the technical specifications, questions were identified for which we require additional information and clarification. Provide responses to the enclosed request for additional information within 60 days from the date of this letter.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.30(b),
Oath or affirmation, you must execute your response in a signed original document under oath or affirmation. Your response must be submitted in accordance with 10 CFR 50.4, Written communications. Information included in your response that is considered sensitive, or proprietary, that you seek to have withheld from the public, must be marked in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding. Any information related to security should be submitted in accordance with 10 CFR 73.21, Protection of Safeguards Information: Performance Requirements. Following receipt of the additional information, we will continue our evaluation of your amendment request.
J. Geuther If you have any questions about this review or if you need additional time to respond to this request, please contact me at 301-415-3965 or via electronic mail at Spyros.Traiforos@nrc.gov.
Sincerely,
/RA/
Spyros A. Traiforos, Project Manager Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-188 License No. R-88
Enclosure:
As stated cc: See next page
Kansas State University Docket No. 50-188 cc:
Office of the Governor State of Kansas 300 SW 10th Avenue, Suite 2125 Topeka, KS 66612-1590 Thomas A. Conley, RRPJ, CHP Section Chief Radiation and Asbestos Control KS Dept of Health & Environment 1000 SW Jackson, Suite 330 Topeka, KS 66612-1365 Mayor of Manhattan City Hall 1101 Poyntz Avenue Manhattan, KS 66502 Test, Research and Training Reactor Newsletter P.O. Box 118300 University of Florida Gainesville, FL 32611
ML17038A272; *concurred via email NRR-088 OFFICE NRR/DPR/PRLB/PM NRR/DPR/PRLB/LA* NRR/DPR/PRLB/BC NRR/DPR/PRLB/PM NAME STraiforos NParker AAdams STraiforos DATE 3/22/17 3/22/17 3/27/17 3/28/17 REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR KANSAS STATE UNIVERSITY NUCLEAR REACTOR FACILITY LICENSE NO. R-88; DOCKET NO. 50-188 The U.S. Nuclear Regulatory Commission (NRC) staff is continuing its review of your license amendment request (LAR), submitted by letter dated April 9, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12109A063), as supplemented by letters dated April 28, 2014, and October 5, 2016 (ADAMS Accession Nos. ML16200A317 and ML16291A498, respectively), to Facility Operating License No. R-88 for the Kansas State University (KSU) training reactor and isotopes production, General Atomics (TRIGA) Mark II Nuclear Reactor (Research Reactor), for the use of up to four fuel elements of 12.5 uranium by weight percent (U wt%) in certain locations of the core of the KSU Research Reactor. This request for additional information (RAI) is based on our review of the LAR, the safety analysis report (SAR) dated December 21, 2004 (ADAMS Accession No. ML052580517), and the KSU technical specifications (TSs), dated March 13, 2008 (ADAMS Accession No. ML080580275).
During our review of this information, questions were identified for which we require additional information and clarification.
The regulatory requirements for a license amendment reside in Title 10 of the Code of Federal Regulation (10 CFR) Section 50.90, Application for amendment of license, construction permit, or early site permit, and states, in part, Whenever a holder of a license desires to amend the license or permit, application for an amendment must be filed with the Commission, as specified in §§ 50.4 of this chapter, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications. The reference to the original applications, as used in the requirements presented in 10 CFR 50.90 refers to the operating license and construction permit applications. The regulations in 10 CFR 50.34, Contents of applications; technical information, provides operating license requirements in 10 CFR 50.34(b). To the extent that this information is also required by the KSU Reactor TS, 10 CFR 50.36, Technical specification, provides those requirements.
NUREG-1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, provides guidance concerning the type and level of detail of the information that is required to support the NRC staff review of your LAR. The information requested is needed by the NRC staff to independently verify that the acceptance criteria discussed in Section 4.6 of NUREG-1537, Part 2, are met by the applicant.
- 1. The regulation in 10 CFR 50.34(b)(2)(i) requires, in part, that an applicant or licensee provide a description and analysis of the structures, system, and components of the facility (in this case the core) with emphasis upon performance requirements, the bases with technical justification upon which such requirements have been established and the evaluations required to show that safety functions will be accomplished. The KSU steady state critical heat flux (CHF) calculation uses a simple pool boiling model that may only be valid for a single rod in a large pool of water. This information is discussed in Section 4.5.3, Fuel and Clad Temperatures, of the SAR dated December 21, 2004.
Enclosure
The model does not account for conditions in the hot channel as is traditionally done by General Atomics (TRD 070.01006.04 Rev. A) and reviewed by Argonne National Laboratory (ANL) (ANL/RERTR/TM-07-01) and gives minimum CHF ratio limits significantly higher than the traditional analyses. Experimental data for TRIGA geometry published in Critical Heat Flux in TRIGA-Fueled Reactors Cooled by Natural Convection, Avery, et al., Nuclear Science and Engineering: 172, 249-258 (2012) shows values of CHF that are much lower than the KSU calculation predicts. Please provide information that validates the pool boiling model used in the SAR or alternatively provide a traditional hot channel analysis using the Bernath correlation (Bernath, L. A Theory of Local Boiling Burnout and Its Application to Existing Data. Chem. Eng. Prog.
Symposium Ser. 56.30 (1960) pp.95-116) with core inlet conditions at the pool TS limits, or a lower temperature validated by facility information that correlates the core inlet conditions with the bulk pool temperature measurement used in the TS limit, or another correlation that has been validated against acceptable data.
- 2. The regulation in 10 CFR 50.34(b)(2)(i) requires, in part, that an applicant or licensee provide a description and analysis of the structures, system, and components of the facility (in this case the nuclear instrumentation system) with emphasis upon performance requirements, the bases with technical justification upon which such requirements have been established and the evaluations required to show that safety functions will be accomplished. Additionally, 10 CFR 50.36(c)(ii)(A), Limiting safety system settings, requires that instrumentation settings are established for automatic protective devices related to variables (reactor power in this case) having a significant safety function (timely actuation). The setting must be chosen such that the automatic protective action will correct the abnormal situation before a safety limit is exceeded.
Based on discussions with the KSU staff, the existing nuclear instrumentation cannot provide power measurements for reactor thermal power levels greater than 1.04 megawatts. The current licensed thermal power limit is 1,250 kilowatts (kWt) and that power level is also identified as a TS limiting safety system setting. As such, the measurement range of the nuclear instrumentation channels are not sufficient to encompass the range necessary for the current licensed thermal power limit. The proposed additional reactivity provided by four 12.5 U wt% fuel elements will likely enable the reactor to achieve thermal power levels that exceed the current measurement range of the existing nuclear instrumentation channels. Nuclear instrumentation is required to be capable to measure the full range of reactor power levels anticipated during normal, transient or accident conditions as described in the safety analyses report. This level of performance is necessary to confirm that design safety functions are or are not successfully accomplished and in the case where the safety function has not been successful, to provide information allowing alternate mitigation and assessment of safety limit exceedance.
Propose a licensed thermal power limit that is within the range of the currently installed nuclear instrumentation or describe how the nuclear instrumentation system is capable of measurement of the full range of reactor power levels anticipated as described in the safety analyses including instrument uncertainties based on the current licensed thermal power limit.
- 3. The thermal-hydraulic and neutronic analyses forms the basis for reactor safety limits, limiting safety system settings and limiting conditions for operation which are required to be included in the facility TS in accordance with 10 CFR 50.36. 10 CFR 50.36(b)
requires that the TS be derived from analyses and evaluations included in the safety analyses report and amendments thereto submitted pursuant to 10 CFR 50.34.
Pursuant to 10 CFR 50.9, Completeness and accuracy of information, requires that information provided to the Commission related to Commission regulations by the licensee shall be complete and accurate in all material respects.
Your original application letter dated April 9, 2012, stated that the core loading at the time of the application could operate at a maximum power of 600 kWt and the revised core loading with four 12.5 U wt% fuel elements in the E or F ring would be able to operate at a power of approximately 900 kWt. In your response to RAI No. 9, dated October 5, 2016, you provided reference to reactor operation on June 27, 2016, at a steady state power level of 735 kWt using the currently licensed core. This is significantly higher than the maximum predicted power of 600 kWt stated in your license amendment application. Given the significance of the deviation of the predicted and actual power observed for the current core causes the NRC staff to question the accuracy of the predicted value of 900 kWt for the core with the four 12.5 U wt% fuel elements. Provide an explanation resolving the significant deviation between the maximum reactor power observed on June 27, 2016 (735 kWt), and the maximum reactor power predicted in support of this license amendment (600 kWt) and provide a revised analysis for the maximum core operating power after the addition of the four 12.5 U wt% fuel elements.
- 4. The accident analyses assumes limiting values for excess reactivity and shutdown margin, which are required by 10 CFR 50.36, as limiting conditions for operation (LCO).
Further, 10 CFR 50.36(b) requires that the technical specifications be derived from analyses and evaluations included in the safety analysis report and amendments thereto submitted pursuant to 10 CFR 50.34. This license amendment application appears to rely on a combination of reactor measurements for excess reactivity and shutdown margin and an estimate of the added reactivity for the 12.5 U wt% fuel elements using the Monte Carlo N-Particle (MCNP) program. Provide clarification on how the MCNP calculations are integrated with the facility measurements to determine the maximum excess reactivity and the minimum shutdown margin.
- 5. The regulation in 10 CFR 50.36(c)(2)(ii) requires that technical specification limiting condition for operation of a nuclear reactor must be established for a structure, system, or component (for example, control rods). Further, 10 CFR 50.36(c)(2)(ii), states a limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria. Criterion 3 states, A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Since control rods meet Criterion 3, appropriate LCO are required to be included in the TS. Additionally, 10 CFR 50.36(b) requires that the technical specifications be derived from analyses and evaluations included in the safety analysis report and amendments thereto submitted pursuant to 10 CFR 50.34.
In a letter dated August 11, 2016, ADAMS Accession No. ML16189A194, the NRC staff provided RAI No. 12 which requested additional information related to TS LCO 3.4 asking for the inclusion of the number of the control rods and the number of operable control rods required for reactor operation. In your response dated October 5, 2016, the requested information and TS revision was not provided. Instead, a justification was
provided as to why the requested information was not necessary. The NRC staff reviewed your response and disagrees with your conclusion that the requested information is not necessary and requests that TS LCO 3.4 be revised to include the number and type of control rods and the minimum number of operable control rods for reactor operation. This information is necessary so that the NRC staff are confident that operation with one or more inoperable control rods fully inserted into the core does not result in the potential for localized fuel damage due to excessive power peaking. If operation with one or more control rods inoperable but fully inserted is acceptable, please provide the supporting analyses and evaluations from which operational acceptability is derived.
- 6. The regulation in 10 CFR 50.36(c) states the categories of information that are required to be included in the TS. Specifically, 10 CFR 50.36(c)(4) Design features, state that Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)
(1), (2), and (3) of this section.
In your original application dated April 9, 2012, it was stated in the Design Constraints, that the 12 U wt% fuel elements may not be placed near control rod channels, to avoid local power peaking effects during pulsing. The NRC staff has reviewed the proposed TS for this license amendment for information relating to safety limitation of the geometric location of the 12.5 U wt% fuel elements in the core. The NRC staff has noted that the information required by 10 CFR 50.36(c)(4) related to the geometric location limitation of the 12.5 U wt% fuel elements has not been included in either the Design Features section of your proposed TS or in 10 CFR 50.36(c)(1), (2), or (3) as appropriate. The NRC staff is also concerned with the usage of the term near as it is ambiguous as to the identification of unacceptable locations.
Provide a revision to the proposed TS describing the geometric limitation and the controls that will help ensure compliance and include information on the acceptability of the specific location where the 12.5 U wt% fuel elements will be installed. Otherwise, describe how the current or previously proposed TS adequately address the control of this geometric limitation on the location 12.5 U wt% fuel elements in the proposed reactor core.