ML19337A435

Univ. of Texas - Austin - Request_ for Extension of Date for Completion of Neutronic and Thermal-Hydraulic Analyses

ML19337A435
Person / Time
Site:	University of Texas at Austin
Issue date:	11/25/2019
From:	Whaley P University of Texas at Austin
To:	Geoffrey Wertz Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2017-RNW-0032
Download: ML19337A435 (2)

Text

DEPARTMENT OF MECHANICAL ENGINEERING Nuciear Engineering Teaching Laboratory Pickle Research Campw R-9000

• Austin, Texas 78758

• 512-232-5380

• FAX 512-471-4589

- nuclear.,mgr.-utexas.edu

• ll'char/ton@austln.utexas.edJI November 25, 2019 ATTN: Docwneni Control Desk U.S. Nuclear Regulatory Commission Washington, D.C._20555-0001 Geoffiey Wert7., P .E.

Non-Power Production and Utilization Facility Licensing Branc~

Division-of Advanced Reactors and Non-Power Utilization Nuclear Reactor Regulation

SUBJECT:

Request_ for Extension of Date for Comp~etion of Neutronic and niermal-Hydraulic Analyses for The University of Texas at Austin

REFERENCE:

October 18, 2018 letter: University of Texas at Austin- Summary of Site Visit and Request for Schedule for Completion of the Reactor Analyses RE: Renewal of Facility Operating License No. R-129 for The University of Texas at Austin Research Reactor (EPID NO. L-2017-RNW-0032)

December 14, 2018 letter, Docket No. 50-602, Facility Operating L i ~ NO. R-129, Schedule for Completion of the Reactor Analyses for the University of Texas at Austin Research Reactor August 12, 2019 letter, Docket No. 50-602, Facility Operating License NO. R-129, Revised Schedule for Completion of the Reactor Analyses for the University of Texas at Austin Research R~tor Sir:

We respectfully request an extension of the scheduled date for providing the neutronics and thermal-hydraulic analyses (supporting relicensing of The University of Texas at Austin nuclear research reactor) by an additional 90-days, to 28 February 2020. -

The Radiation Center of the Oregon State University (OSU) developed a neutronics model_ us_ing MCNP to support safety analysis, and provided preliminary results. The results were reviewed with the Oregon State University analyst Calculations of excess reactivity are within a reasonable range for the OSU analysi$, but the calculated control rod worth values are ~ot. Potential for improvement w(ls identified.

Specification of the UT control rods is based on General Atomics schematics at the facility. The UT TRIGA core has a hexagonal pitch while most U.S. TRIGA reactors are either circular or rectangular geometry, not directly comparable. The Moroccan TRIGA design .is a hex-core, similar to the UT--_

TRIGA. There are a number of publications apalyzing the Moroccan reactor in print or in press based on MCNP. Comparison of the UT and Moroccan MCNP models could either identify potential sources of m

error in the UT model or provide a level of confidence the UT model. A request has been made for a copy of the input file from the Moroccan facility.

A single nominal value for fuel density was used in the UT TRIGA MCNP model, but the fuel density for a

the elements used in the initial core are neither nominal nor single valued. Using value for fuel density consistent with_ the uranium mass and enrichment for the individual elements could remove a potential source of error. -*

DEPARTMENT OF MECHANICAL ENGINEERING, Nuclear Engineering Teaching Laboratory f'ickle Research Catnpu3 R-9000

• Austin, Texas 78758

• 5 l 2-232-5380

• F,1.X 512-471-4589

. mtelear.engr.utexru.edu*wchar/ton@austin.utexas.edu Th~ iso~opic ~ncentrations used in the initial ~re were taken from a OT TRIGA model previously developed in SCALE (T-6 depletion sequence). There are significant differences between SCALE depletion and depletion in MCNP and, although the composition may prove comparable, using MCNP to derive the initial composition of the fuel elements with prior operating history (i.e., all fuel elements except for those in the fuel follower control rod) could remove a potential source of error.

• ne MCNP model developed by OSU included three segments for ~h fuel element Ho:wever, because of the complication involved in mapping the variety of serial numbers' at UT into a manageable scheme only one fuel material was used for each fuel element Fuel followers are not completely exposed in the core during reactor operation and, as a minimum, the fuel follower materials could be segmented and burned separately to reduce a potential source of error.

• Other work {comp~eted or in progress) based on the discussion to improve the analysis includes:

The MCNP model developed by OSU was generated using arrays and three segments for each fuel.

element A revised model without arrays returned comparable results, indicating array methodology does not introduce errors.

The MCNP model developed by OSU used natural borori cross section libraries to characterize the boron carbide poison section (?fthe control rods, but the latest version ofENDF defined natural boron as graphite rather than (preYious) boron isotopes in natural concentrations. The carbon in boron carbide is not graphite. Following confirmation with analysts at-Idaho National Laboratories, the boron specification was changed to the naturally occurring isotopes with a minor effect on criticality calculations.

The Moroccan MCNP model developer suggests the criticaiity simulations for control rod worth should be perfonned in a manner that reflects the actual measurement more closely. OSU is currently performing analysis to determine if this is a potential source of error.

We have identified other potential diagnostic operations as well.

As note, OSU is currently performing calculations and I am hopeful ~t we will be able to see the Moro~ model. The remaining work is not trivial and coordination with the holiday will ~ required; the OSU analyst indicates the work can be completed by February 28.

,,.""'""'--"Y questions, please contact me at 512-232-5373 or whaley@mail.utexas.edu I declare under penalty of perjury that the foregoing is true and correct

• I/hr/~

W. S. Charlton