ML13133A349

University of Missouri - Columbia, Amendment Reactor Core Safety Limits

ML13133A349
Person / Time
Site:	University of Missouri-Columbia
Issue date:	07/08/2013
From:	Alexander Adams Research and Test Reactors Licensing Branch
To:	Rhonda Butler Univ of Missouri - Columbia
Adams, A
References
TAC ME7018
Download: ML13133A349 (24)

Text

July 8, 2013 Mr. Ralph Butler, Director Research Reactor Center University of MissouriColumbia Research Park Columbia, MO 65211

SUBJECT:

UNIVERSITY OF MISSOURICOLUMBIA AMENDMENT RE: REACTOR CORE SAFETY LIMITS (TAC NO. ME7018)

Dear Mr. Butler:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 36 to Amended Facility License No. R-103 for the University of Missouri-Columbia Research Reactor.

The amendment consists of changes to Section 2.1 of the technical specifications, Reactor Core Safety Limits in response to your application of August 24, 2011, as supplemented on May 23 and July 30, 2012.

A copy of the safety evaluation supporting Amendment No. 36 is enclosed.

Sincerely,

/RA/

Alexander Adams, Jr., Chief Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-186

Enclosures:

1. Amendment No. 36

2. Safety Evaluation cc w/enclosures:

See next page

July 8, 2013 Mr. Ralph Butler, Director Research Reactor Center University of MissouriColumbia Research Park Columbia, MO 65211

SUBJECT:

UNIVERSITY OF MISSOURICOLUMBIA AMENDMENT RE: REACTOR CORE SAFETY LIMITS (TAC NO. ME7018)

Dear Mr. Butler:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 36 to Amended Facility License No. R-103 for the University of Missouri-Columbia Research Reactor.

The amendment consists of changes to Section 2.1 of the technical specifications, Reactor Core Safety Limits in response to your application of August 24, 2011, as supplemented on May 23 and July 30, 2012.

A copy of the safety evaluation supporting Amendment No. 36 is enclosed.

Sincerely,

/RA/

Alexander Adams, Jr., Chief Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-186

Enclosures:

1. Amendment No. 36

2. Safety Evaluation cc w/enclosures:

See next page DISTRIBUTION:

PUBLIC RidsNrrDpr RidsNrrDprPrta RidsNrrDprPrtb ADAMS Accession No: ML13133A349 OFFICE PRLB:LA PRLB:BC OGC PPR:DD PRLB:BC NAME CHawes AAdams SUttal MMuessle AAdams DATE 5/29/13 5/30/13 6/7/13 7/8/13 7/8/13 Official Record Copy

University of Missouri-Columbia Docket No. 50-186 cc:

John Ernst, Associate Director Regulatory Assurance Group Research Reactor Facility Columbia, MO 65201 Homeland Security Coordinator Missouri Office of Homeland Security P.O. Box 749 Jefferson City, MO 65102 Planner, Dept of Health and Senior Services Section for Environmental Public Health 930 Wildwood Drive, P.O. Box 570 Jefferson City, MO 65102-0570 Deputy Director for Policy Department of Natural Resources 1101 Riverside Drive Fourth Floor East Jefferson City, MO 65101 A-95 Coordinator Division of Planning Office of Administration P.O. Box 809, State Capitol Building Jefferson City, MO 65101 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

UNIVERSITY OF MISSOURI-COLUMBIA DOCKET NO. 50-186 AMENDMENT TO AMENDED FACILITY LICENSE Amendment No. 36 License No. R-103

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. The application for an amendment to Amended Facility License No. R-103, filed by the University of Missouri-Columbia (the licensee) on August 24, 2011, as supplemented on May 23 and July 30, 2012, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended, and the regulations of the Commission as stated in Title 10, Chapter I, Nuclear Regulatory Commission, of the Code of Federal Regulations (10 CFR Chapter I).

B. The facility will operate in conformity with the application, the provisions of the Atomic Energy Act of 1954, and the rules and regulations of the Commission.

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

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

E. This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, and the licensee has satisfied all applicable requirements.

F. Prior notice of this amendment was not required by 10 CFR 2.105, Notice of Proposed Action, and publication of a notice for this amendment is not required by 10 CFR 2.106, Notice of Issuance.

2. Accordingly, the license is amended by changes to the technical specifications as indicated in the enclosure to this license amendment, and paragraph 3.B of Amended Facility License No. R-103 is hereby amended as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 36, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Mary C. Muessle, Deputy Director Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Enclosures:

1. Amended Facility License No. R-103

2. Changes to Appendix A, Technical Specifications Date of Issuance: July 8, 2013

ENCLOSURE 1 TO LICENSE AMENDMENT NO. 36 AMENDED FACILITY LICENSE NO. R-103 DOCKET NO. 50-186 Replace the following page of Amended Facility License No. R-103 with the enclosed page.

The revised page is identified by amendment number and contains a vertical line indicating the area of change.

Remove Insert 3 3

A. Maximum Power Level The licensee may operate the reactor at steady state power levels up to a maximum of 10 MWt.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 36, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Physical Security Plan The licensee shall maintain and fully implement all provisions of the Commissions approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 73.21 entitled Physical Security Plan for University of Missouri Research Reactor Facility dated June 10, 1983, submitted by letter dated June 10, 1983.

4. This amended license is effective as of date of issuance and shall expire at midnight on October 11, 2006.

FOR THE ATOMIC ENERGY COMMISSION

/RA/

Karl R. Goller Assistant Director for Operating Reactors Directorate of Licensing

Attachment:

Appendix A (Change No. 10 to the Technical Specifications)

Date of Issuance: July 9, 1974 Amendment No. 36 July 8, 2013

ENCLOSURE 2 TO LICENSE AMENDMENT NO. 36 AMENDED FACILITY LICENSE NO. R-103 DOCKET NO. 50-186 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines that indicate the areas of change.

Remove Insert 2.1 pages 1 to 6 2.1 pages 1 to 6

TECHNICAL SPECIFICATION UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY Number 2.1 Page 1 of 6 Date July 8, 2013

SUBJECT:

Reactor Core Safety Limit Applicability This specification applies to reactor power and reactor coolant system flow, temperature and pressure.

Objective The objective is to set forth parameter safety limits which shall prevent damage to the fuel element cladding.

Specification Reactor power, coolant system flow, temperature and pressure shall not exceed the following limits during reactor operation.

a. Mode I and II (Core Flow Rates 400 gpm)

The combination of the true values of the reactor power level, core flow rate, and reactor inlet water temperature shall not exceed the limits described by Figures 2.0, 2.1, and 2.2. The limits are considered exceeded if, for flow rates greater than 400 gpm, the point defined by the reactor power level and core flow rate is at any time above the curve corresponding to the true values of the reactor inlet water temperature and primary coolant system pressurizer pressure. To define values of the safety limits for Amendment No. 36

Number 2.1 Page 2 of 6 Date July 8, 2013 Amendment No. 36

Number 2.1 Page 3 of 6 Date July 8, 2013 Amendment No. 36

Number 2.1 Page 4 of 6 Date July 8, 2013 Amendment No. 36

TECHNICAL SPECIFICATION UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY Number 2.1 Page 5 of 6 Date July 8, 2013

SUBJECT:

Reactor Core Safety Limit (continued) temperatures and/or pressures not shown in Figures 2.0, 2.1, and 2.2, interpolation or extrapolation of the data on the curves shall be used.

For pressurizer pressures greater than 85 psia, the 85 psia curves (Figure 2.2) shall be used and no pressure extrapolation shall be permitted.

b. Mode I and II (Core Flow Rates < 400 gpm)

Steady state power operations in Modes I and II are not authorized for a core flow rate < 400 gpm. Reactor operations with core flow below 400 gpm will occur only after a normal reactor shutdown when the primary coolant pumps are secured or following a loss of flow transient. Under the above conditions the maximum fuel cladding temperature shall not exceed 366 oF.

c. Mode III Reactor Power ......................................................... 150 Kilowatts (maximum)

Bases

a. A complete safety limit analysis for the MURR is presented in Appendix F of Addendum 4 to the Hazards Summary Report (HSR). A family of curves is presented which relate reactor inlet water temperature and core flow rate to the reactor power level corresponding to a Critical Heat Flux (CHF) ratio of 2.0 based on the Bernath CHF Correlation. This also corresponds to a flow instability Departure from Nucleate Boiling Ratio (DNBR) of 1.2 based on the burnout heat flux data experimentally verified for ATR type fuel elements. Curves are presented for pressurizer pressures of 60, 75, and 85 psia. The safety limits were Amendment No. 36

TECHNICAL SPECIFICATION UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY Number 2.1 Page 6 of 6 Date July 8, 2013

SUBJECT:

Reactor Core Safety Limit (continued) chosen from the results of this analysis for Mode I and II operation, i.e. forced convection operation above 400 gpm flow.

b. Steady state reactor operation is prohibited for core flow rates below 400 gpm by the low flow scram settings in the safety system. The region below 400 gpm will only be entered following a reactor shutdown when the primary coolant pumps are secured or during a loss of flow transient where the reactor scrams, the flow coasts down to zero, reverses, and natural convection cooling is established. Below 400 gpm core flow the criterion for the safety limit is that fuel plate temperature must be less than 900 °F; the temperature at which fuel cladding failure could occur.

The analysis of a loss of flow transient from the ultra-conservative conditions of 11 MW of power, 3000 gpm core flow and 155 °F core inlet temperature indicated a maximum fuel cladding temperature of 327 °F which is well below the cladding DNB temperature of 366 °F.

c. Analysis of natural convection cooling of the core (Mode III operation) is presented in section 5.5.3 of the HSR.

Amendment No. 36

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 36 TO AMENDED FACILITY LICENSE NO. R-103 THE UNIVERSITY OF MISSOURI-COLUMBIA DOCKET NO. 50-186

1.0 INTRODUCTION

By letter dated August 24, 2011, as supplemented on May 23 and July 30, 2012 (Agencywide Document Access and Management System (ADAMS) accession numbers ML11237A088, ML12150A050, ML12150A052 and ML12214A310), the University of Missouri-Columbia (the licensee) requested a license amendment to change Appendix A of Amended Facility License No R-103, Technical Specifications for University of Missouri Research Reactor Facility. The amendment would revise technical specification (TS) 2.1, Reactor Core Safety Limit.

2.0 BACKGROUND

The licensee operates the Missouri University Research Reactor (MURR) at its campus in Columbia, Missouri. The MURR is licensed to operate in three modes of operation:

Mode I not to exceed a thermal power level limit of 10 megawatts (MW(t)), Mode II not to exceed 5 MW(t), and mode III not to exceed a thermal power limit of 50 kilowatts (kW(t))

(Mode III is limited to operation with natural convection cooling).

The regulations in Title 10 of the Code of Federal Regulations, Part 50, Section 36 (10 CFR 50.36) require licensees to have TSs. The regulations at 10 CFR 50.36(c)(1) requires, in part, that TSs include safety limits. Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. In the case of the MURR, the primary physical barrier is the fuel cladding. The reactor is designed to operate under a set of safety limit curves developed to ensure that flow instability and departure from nucleate boiling (DNB) are prohibited. This helps to ensure that fuel melt or fuel clad damage is prevented. The safety limit curves are based on three measureable independent operating parameters:

primary coolant flow rate, reactor inlet water temperature, and pressurizer pressure.

These three operating parameters determine the reactor power level safety limit. The safety limit curves provide the basis for the Mode I and Mode II operating limits and the Limiting Safety System Settings (LSSS) set points for reactor safety system activation for each operating mode. The Mode III operating limit is only based on reactor power level.

The reactor has two primary cooling loops. Operating Mode I has both cooling loops in operation while Mode II only requires one cooling loop. TS 2.2 lists the LSSS set points

for Mode I or II as 125 percent of power, 155 degrees F reactor inlet temperature, 1625 gpm coolant flow rate for loop flow (two loops in Mode I for a total flow of 3250 gpm and one loop in Mode II for 1625 gpm), and 75 psia at the pressurizer. The normal operating conditions for Mode I (10 MW(t)) operation are reported by the licensee (response to Request for Additional Information (RAI) 4.18b, ADAMS Accession No. ML12355A019) as 3800 gpm coolant flow rate, 120 degrees F inlet temperature, and greater than 75 psia coolant system pressure at the pressurizer. Using the worst case LSSS set points and allowing for instrument uncertainty, a power safety limit of about 14.75 MW(t) was calculated (Response to RAI 4.18c, ADAMS Accession No. ML12355A019). The amount of margin between the LSSS and the safety limit is 2.25 MW(t). The licensee concluded that this value is sufficiently large to accommodate measurement uncertainty of operating parameters.

On January 17, 2011, the licensee submitted a written communication (ADAMS Accession No. ML1101906010) to the U. S. Nuclear Regulatory Commission (NRC) as required by MURR TS 6.1.h(2) stating that an error was found in the Hazards Summary Report (HSR) which dated back to 1973. The error was found while answering RAI question 16.1 (ADAMS Accession No. ML101160244) posed by the NRC staff as part of the relicensing review of the MURR facility.

The Bernath correlation is used to determine the critical heat flux in the limiting reactor cooling channel. The licensee explained that the error was caused by misinterpreting the definition of diameter of heated surface (Di) in the Bernath correlation as heated diameter (Dh). For water flowing in a heated tube, the definitions are the same.

However, the MURR flow cross section is a thin rectangular channel heated along the two long sides, which produces an order of magnitude difference between the two definitions.

As discussed in letter dated January 17, 2011, the licensee applied the correction to the methodology for calculating the safety limit parametric curves. The licensee also used updated peaking factors in the calculations based on modern calculation models. The result of these calculations was that lowering the reactor inlet temperature LSSS from 155 degrees F to 153 degrees F maintains about the same margin between the safety limit and the LSSS of TS 2.2. Using the new temperature LSSS of 153 degrees F, the licensee determined the SL to be 14.955 MW(t), a margin of 2.45 MW(t). The licensee also performed the safety limit calculations with just the correction to the Bernath correlation without using updated peaking factors. This calculation showed a margin of 1.24MW(t) for a temperature of 150 degrees F. The licensee operates the reactor with a reactor inlet temperature limit scram set point of 148 degrees F (the 150 degree F temperature represents the worse case error of 2 degrees F on the 148 degree F scram set point limit) which is within the 153 degree F LSSS. The normal operating reactor inlet temperature is about 120 degrees F. The NRC staff reviewed this information and agreed that no immediate actions were needed and this error could be corrected by a license amendment.

The August 24, 2011, license amendment request submitted by MURR requested changes to TS 2.1, Reactor Core Safety Limit, Figures 2.0, 2.1, and 2.2, to correct the error identified in the January 17, 2011, report. The licensee identified that the error also exists in the MURR HSR Appendix F of Addendum 4 base document Safety Limit Analysis for the MURR facility developed by a licensee subcontractor while answering license renewal RAI question 4.17 (ADAMS accession No. ML101160266). Information

related to the license renewal will be revised after issuance of this license amendment.

The licensee submittal proposes new power peaking factors based on refined analyses discussed above. The licensee has determined that the new power peaking factors permit the reactor inlet temperature to remain at 155 degrees F instead of the 153 degrees F discussed in the report of January 17, 2011. Raising the reactor inlet temperature from 153 to 155 degrees F lowers the power safety limit from 14.955 to 14.894 MW(t). The licensee has only requested changes to TS 2.1. The reactor LSSSs are not changed.

A RAI was sent to the licensee on April 12, 2012 (ADAMS Accession No. ML121010109). The licensee provided additional information regarding their amendment request in correspondence dated May 23, 2012 (ADAMS Accession No. ML12150A050), and July 30, 2012 (ADAMS Accession No. ML12214A310). The May 23, 2012, letter requested other administrative changes to the TSs and TS bases changes to TS 2.1.

3.0 EVALUATION 3.1 Safety Limit Curves The licensee has requested that the safety limit curves given in figures 2.0, 2.1 and 2.2 of TS 2.1 be revised to correct the error caused by misinterpreting the definition of diameter of heated surface (Di) in the Bernath correlation as heated diameter (Dh) as discussed above. It was also requested that the safety limit curves be revised to use peaking factors based on modern analysis.

The NRC staff reviewed the licensees previous safety limit analysis in light of the licensee-identified error to understand the initial conditions and assumptions that were used in that analysis. Based on this review and a review of the Bernath correlation, the NRC staff finds that the licensee has correctly identified the error in the parameter used in the previous safety analysis. This error affects the safety limit curves in TS figures 2.0, 2.1, and 2.2 of TS 2.1.

To evaluate the revised safety limit curves proposed by the licensee, the NRC staff reviewed the license amendment request, as supplemented, and the relevant sections of the facility Safety Analysis Report and TS. The NRC staff reviewed the licensees derivation of the safety limit curves and the assumptions used in the derivation. The following assumptions continued from the original analysis:

• The licensee assumes that the channel flow area does not vary with axial position and coolant velocities in all channels are equal. The justification is that manufacture tolerances and uncertainties have been included in the hot channel factors for flow and enthalpy rise. The NRC staff has reviewed the licensees methodology in determining this hot channel factor and has finds it acceptable.

• The licensee assumes that bulk coolant in flow channels is always sub-cooled. The justification is that bulk boiling at the channel exit is not permitted. The NRC staff has reviewed the licensees methodology used in the spreadsheet calculation presented in the amendment request and has found it to be an acceptable method in determining the maximum reactor power level with a DNB ratio (DNBR) of 1.2 at the

hot channel exit. This value for the DNBR had been found acceptable by the NRC staff in the original licensing base for the MURR.

• The licensee assumes that 93 percent of energy generated is deposited in the fuel plate. The justification for this assumption is that 7 percent of fission power escapes the fuel by gamma or neutron emission. The NRC staff has reviewed and compared this assumption to other published sources and has found this value is more conservative than published energy deposition validation studies at the Advanced Test Reactor (ATR) which has a similar fuel design as the MURR.

• The licensee assumes that flow instability occurs at 50 percent of value predicted by the Bernath correlation. The justification for this assumption is that testing at the ATR has shown flow instability induced DNB occurs at 60 percent of the value predicted by the Bernath correlation. Thus, the use of 50 percent of value predicted by the Bernath correlation adds further margin to DNB. This results in a DNBR of 1.2 based on the ratio of 0.6 to 0.5. The NRC staff has reviewed the reference materials and finds that the flow instability assumption is appropriate.

• The licensee assumes turbulent flow (Blasius equation) for frictional pressure drop in components (core, pipes, pipe elbows, and valves). The justification is that the Reynolds number values are in the turbulent region under all flow ranges for MODE I and MODE II operation. The NRC staff has reviewed the flow data and finds that the pressure drop assumptions are appropriate for the flow conditions.

• The licensee assumes a reactor power of 10 MW(t) for heat flux power peaking factors. The justification for this assumption is that 10 MW(t) is the licensed power level for the MURR. Further, the heat flux profile used in the analysis is normalized.

As a result, assuming a higher reactor power would not change the location of the hot channel. The NRC staff reviewed the methodology for the determination of peaking factors and finds that the conclusion that the location of the hot channel is independent of reactor power is well founded.

• The licensee assumes that the difference in fuel meat (fueled part of the fuel plate) arc lengths between plates can be ignored. The effect of this assumption was shown in the licensees amendment documentation to be negligible. The NRC staff has reviewed that analysis and finds that this assumption is reasonable.

• The licensee assumes that the difference in coolant density between hot and nominal channels can be ignored. The effect of this assumption was shown in the licensees amendment documentation to be negligible. The NRC staff has reviewed that analysis and based on the small differences in temperature and the relative insensitivity of coolant density to the small temperature difference finds that this assumption is reasonable.

The NRC staff finds the justification for the unchanged assumptions and initial conditions discussed above to be still valid and consistent with previously approved methodology.

The licensee updated some initial conditions and assumptions from those in the original analysis to reflect advancements in calculation methods. The NRC staff has evaluated these changes as discussed below.

• The licensee applies the exit pressure to all axial positions in the core when calculating the DNBR. In the previous analysis, the licensee used the pressurizer pressure throughout the primary loop. The NRC staff finds this use of the lower core exit pressure to be conservative in calculating the DNBR as it will result in a smaller margin to the LSSS and lowers the allowable safety limit power.

• The licensee included a small variation of bulk coolant density within the channel in the model. The original analysis did not include this variation. This change in coolant density increases the core coolant velocity at the core exit location. This increase in coolant velocity increases the calculated pressure loss across the core region and thus lowers the core exit pressure and the reactor power safety limit. The NRC staff reviewed the modeling of coolant density and finds the conclusions appropriate and conservative as they result in a lower safety limit value.

• The licensee has calculated updated values for hot channel factors for both heat flux and enthalpy rise in the new analysis. Notably, the axial factor was lowered from 1.432 in the prior analysis to 1.2958. In their response to RAIs (ADAMS Accession No. ML12150A052) from the NRC staff regarding this lower axial factor, the licensee explained that the original analysis was performed using a two dimensional core model and described in more detail the model that was used in the new analysis.

The licensee also stated that two assumptions were used in the original model regarding control blade position and burnup that while resulting in a simplified conservative calculation of hot channel factors could not simultaneously occur in actual operation of the MURR. For the new analysis, the licensee used a three dimensional Monte Carlo N-Particle (MCNP) transport code model, which is a computer core analysis program that was not available at the time of the original analysis. The licensee also adjusted the location of the limiting power peaking position based on the results of their calculations. The new location used for the determination of core power peaking as determined by analysis was limiting for safety limit for all but six of 180 data points in the core model. The licensee stated that those six data points with more limiting peaking factors were restricted by saturation temperature at the coolant channel exit and other flow considerations.

The licensee concluded that the new limiting power peaking location was appropriately chosen.

The NRC staff reviewed the licensees modeling techniques and results and finds that the use of the MCNP code for three dimensional analysis of the MURR core is an appropriate technique and has been used for analysis of this type at numerous other reactor and non-reactor facilities. The NRC staff finds that the licensees use of the 1.2958 axial peaking factor is consistent with the additional precision gained from the modeling in three rather than two dimensions along with the revised initial conditions regarding fuel burnup and rod positioning.

• In a RAI, the NRC staff asked about the use of an additional peaking factor of 1.062 which was not present in the original analysis. The licensee responded that the new MCNP code analysis resulted in additional margin that could have justified increasing the reactor safety limit. To keep the licensed safety limits at about the same value, an additional peaking factor adjustment of 1.062 was used. The resulting new safety

limit curves differ slightly from the previous safety limit curves based on the new calculations.

The NRC staff reviewed the licensees calculations and stated rationale and finds that the use of the additional peaking factor of 1.062 is conservative and increases the margin of safety of the reactor.

• The licensee in their original analysis uses a normalized axial power distribution for the hot channel based on beginning of core life with control blades half-in.

Therefore, the most limiting flux is at the core exit (Figure 1 from MURR HSR, Addendum 4, Appendix F). The new analysis uses heat flux (power distribution) taken from the MCNP code analysis results of the week 58 MURR fuel cycle from TDR-0125, Feasibility Analyses for HEU to LEU Fuel Conversion of the University of Missouri Research Reactor (MURR), with no xenon and the flux trap region containing 100 percent water.

The NRC staff reviewed the bases for the calculation approach and finds that the use of the 58 week burnup core instead of a fresh core provides higher peaking factors due to the large heat flux differences between assemblies which is acceptable and conservative.

• The new licensee analysis was modeled based on the presence of only water and no experiments in the central flux trap region of the core. An RAI from the NRC staff asked the licensee for justification of this modeling assumption. The licensee responded that the most reactive condition for the central flux trap region of the core is with all water and that any other materials lower the peaking factor in the adjacent core. The licensee presented results calculated for modeling cases including fueled experiments (to the limit allowed by the facility license) and for materials that have a greater moderating effect than water.

Based on its review, the NRC staff agrees with the licensees calculations which support the conclusion that the calculated safety limit for the reactor based on an all-water flux trap region of the core is limiting.

• The licensees modeling accounted for azimuthal heating of each plate by use of an azimuthal peaking factor. This was justified as the added peaking factor adjusts the model to include azimuthal hot stripes in a channel instead of an average heat flux across the arc length of the fuel plate. No credit was taken for azimuthal mixing in the cooling channel or azimuthal heat conduction. The channel hot stripes are located directly across the cooling channel from each other. The NRC staff reviewed the use of azimuthal peaking factors by the licensee and finds the methodology used to be conservative.

• The licensee used measured data and a benchmarked model for the pressure drop from the pressurizer to the core. Further, the pressure drop model includes gravity head in the core exit pressure and uses coolant velocity at core exit rather than in the main loop. This model determines the pressure at different locations throughout the primary system rather than imprecisely using a single pressure value for all points in the reactor.

The NRC staff reviewed the pressure drop modeling and finds that the use of measured data and a benchmarked model to more precisely calculate the differential pressure from the pressurizer to the reactor core inlet is appropriate. The NRC staff further finds that the addition of the gravity head is conservative as it reduces the core exit pressure and results in lower values for the reactor power safety limit.

Finally, the NRC staff finds that the use of the higher coolant core exit velocities increases the pressure loss and lowers the resulting calculated reactor power safety limit. As the resulting power safety limit is lower, the NRC staff finds it to be conservative and therefore, acceptable.

The NRC staff has reviewed the calculation model used to derive the revised safety limit curves. Based in its review, the NRC staff finds the justification for the modified assumptions and initial conditions to be appropriate and conservative. The above discussion of calculation model assumptions and initial conditions captures the significant aspects of the calculation model. The NRC staff finds that the licensee used an acceptable calculation model to derive the revised safety limit curves.

The licensee has proposed changes to the bases of TS 2.1.a. A sentence that states An extension of this analysis is presented in Section 6 of Addendum 5 to the HSR. is deleted. This reflects the fact that the revised Appendix F of Addendum 4 to the Hazards Summary Report provided in the amendment application replaces the existing Section 3.3 of Addendum 3, Appendix F of Addendum 4 and Section 6.0 of Addendum 5 to the Hazards Summary Report. Because the proposed change updates the references to the Hazards Summary Report in the TS bases to reflect the requested license amendment, the change is acceptable to the NRC staff.

The Bases for TS 2.1.a. reads in part:

A family of curves is presented which relate the reactor inlet water temperature and core flow rate to the reactor power level corresponding to a DNB ratio (DNBR) of 1.2 based on burnout heat flux data experimentally varified for ATR type fuel elements.

To reflect the requested license amendment, the licensee has proposed replacing this sentence with:

A family of curves is presented which relate reactor inlet water temperature and core flow rate to the reactor power level corresponding to a Critical Heat Flux (CHF) ratio of 2.0 based on the Bernath CHF Correlation. This also corresponds to a flow instability Departure from Nucleate Boiling Ratio (DNBR) of 1.2 based on the burnout heat flux data experimentally verified for ATR type fuel elements.

Because the changes update the bases to reflect the requested license amendment, the changes are acceptable to the NRC staff.

The NRC staff has reviewed the proposed revisions to the MURR safety limit curves.

Based on the above findings, the NRC staff concludes that the methodology and assumptions used by the licensee to determine the thermal-hydraulic operating envelope used to establish the safety limit curves in the revised Figures 2.0, 2.1, and 2.2 are technically correct and appropriate for the MURR facility. Further, the NRC staff

concludes that the MURR can be operated using these revised safety limit curves with no undue risk to the health and safety of the public or the environment.

3.2 Administrative Changes to TS 2.1 The licensee has proposed a number of changes to improve and clarify the wording of TS 2.1.

The licensee has proposed changes to the TS to improve grammar. The applicability section of the TS reads as follows:

This specification applies to reactor power and reactor coolant system flow temperature and pressure.

The licensee has proposed adding a comma after the word flow to read as follows:

This specification applies to reactor power and reactor coolant system flow, temperature and pressure.

TS 2.1.a. reads in part:

For pressurizer pressures greater than 85 psia the 85 psia curves (Figure 2.2) shall be used and no pressure extrapolation shall be permitted.

The licensee has proposed adding a comma between 85 psia and the 85 psia to read as follows:

For pressurizer pressures greater than 85 psia, the 85 psia curves (Figure 2.2) shall be used and no pressure extrapolation shall be permitted.

TS 2.1.b. reads in part:

Steady state power operations in Modes I and II is not authorized for a core flow rate < 400 gpm.

The licensee has proposed changing is to are to read as follows:

Steady state power operations in Modes I and II are not authorized for a core flow rate < 400 gpm.

The bases for TS 2.1.a. reads in part:

The safety limits were chosen from the results of this analysis for Mode I and II operation, i.e., forced convection operation above 400 gpm flow.

The licensee has proposed removing the comma after i.e. to read as follows:

The safety limits were chosen from the results of this analysis for Mode I and II operation, i.e. forced convection operation above 400 gpm flow.

The NRC staff has reviewed these proposed changes are finds that they are grammatical in nature and do not change the meaning of the TS. Therefore, the NRC staff concludes these changes are acceptable.

TS 2.1.a. reads in part:

The combination of the true values of the reactor power level, core flow rate, and reactor inlet temperature shall not exceed the limits described by Figures 2.0, 2.1, and 2.2. The limits are considered exceeded if, for flow rates greater than 400 gpm, the point defined by the reactor power level and core flow rate is at any time above the curve corresponding to the true values of the reactor inlet temperature and primary coolant system pressurizer pressure.

The licensee has proposed changing reactor inlet temperature to reactor inlet water temperature at two places. The licensee states the purpose of the change is to remove ambiguity. The change also matches wording on the safety limit curves.

The combination of the true values of the reactor power level, core flow rate, and reactor inlet water temperature shall not exceed the limits described by Figures 2.0, 2.1, and 2.2. The limits are considered exceeded if, for flow rates greater than 400 gpm, the point defined by the reactor power level and core flow rate is at any time above the curve corresponding to the true values of the reactor inlet water temperature and primary coolant system pressurizer pressure.

Because this change clarifies the TS it is acceptable to the NRC staff.

The licensee has proposed rotating the safety limit curves in Figures 2.0, 2.1 and 2.2 by 90 degrees to improve functionality.

Because this change improves the functionality of the figures it is acceptable to the NRC staff.

The licensee has proposed a change to the Bases for TS 2.1.b. which is not directly related to the license amendment request. The Bases for TS 2.1.b. reads in part:

Below 400 gpm core flow the criterion for the safety limit is that the fuel plate temperature must be below that temperature which would result in fuel cladding failure.

The licensee has proposed updating this sentence to read as follows:

Below 400 gpm core flow the criterion for the safety limit is that fuel plate temperature must be less than 900 °F; the temperature at which fuel cladding failure could occur.

The licensee explained that during fabrication all MURR fuel assemblies undergo a cladding bond integrity test that includes blister testing by heating the plates to 900 degree F for a period of two hours. This testing is a requirement of the fuel plate

fabrication specifications. Because the requested changes more accurately describe the basis of the limits applied to the reactor, the changes are acceptable to the staff.

The NRC staff has reviewed these proposed changes and finds that they clarify the TSs without changing the meaning of the TS. Therefore, the NRC staff concludes that these changes are acceptable.

4.0 ENVIRONMENTAL CONSIDERATION

Section 3.1 of this amendment involves changes in the installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20. The licensee revised the safety limits curves to correct the error that was made in the application of the Bernath correlation and to use modern calculation methods. The results are new safety limit curves similar to the old curves. No changes to the LSSSs set points were needed and the safety margins between the safety limits and LSSSs set point are not significantly changed. The amendment does not change the operation of the MURR. For these reasons there is no significant increase in the amounts or change in the types of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure.

Operation of the MURR in accordance with the proposed license amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated because the safety limits and LSSSs set points are not significantly changed. For this same reason, operation of the MURR in accordance with the proposed license amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated or involve a significant reduction in a margin of safety. For these reasons the NRC staff concludes that the proposed amendment involves no significant hazards consideration.

The staff has determined that this amendment involves no significant hazards consideration, no significant increase in the amounts or change in the types of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion as set forth in 10 CFR 51.22(c)(9).

Section 3.2 of this amendment involves changes to the format of the license or otherwise makes editorial, corrective or other minor revisions. Accordingly, this amendment meets the eligibility criteria for categorical exclusion as set forth in 10 CFR 51.22(c)(10)(v).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

5.0 CONCLUSION

S The staff has concluded, on the basis of the considerations discussed above, that (1) the amendment does not involve a significant hazards consideration because the amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, create the possibility of a new kind of accident or a different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such

activities will be conducted in compliance with the Commission's regulations, and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public.

Principal Contributors: A. Adams, Jr., NRC J. Willison, URS Safety Management Solutions C. Neill, URS Safety Management Solutions Date: July 8, 2013