ML12100A097

U.S. Geological Survey Triga Reactor, Response to Request for Additional Information to Question 14

ML12100A097
Person / Time
Site:	U.S. Geological Survey
Issue date:	03/28/2012
From:	Timothy Debey US Dept of Interior, Geological Survey (USGS)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML12100A097 (3)

Text

SUSGS science for a changing world Department of the Interior US Geological Survey PO Box 25046 MS 974 Denver, CO 80225-0046 March 28, 2012 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Reference:

U.S. Geological Survey TRIGA Reactor (GSTR), Docket 50-274, License R-1 13 Request for Additional Information (RAI) dated September 29, 2010

Subject:

Response to Question 14 of the Referenced RAI Mr. Wertz:

Question 14:

14.1 In GSTR SAR Section 13.1, the G STR SAR states that fuel temperature limits of 1,100 'C for stainless steel clad fuel and 535 °C for aluminum clad fuel were set to preclude the loss of clad integrity. GSTR SAR Technical Specification (TS)

Section 14.2.1 lists safety limits for stainless steel clad fuel and aluminum fuel of 1,000 'C and 535 'C, respectively. The NRC has accepted (NUREG-1537 Appendix 14.1) that no fuel damage or cladding failure is expected ifthe fuel temperature for aluminum clad fuel is maintained at a temperature of less than 500 °C and the stainless steel clad fuel temperature is not to exceed 1,150 0C when the cladding temperature is less than 500 0C and is not to exceed 950 'C if the cladding temperature is greater than 500 °C. Please provide justification for the use of fuel temperature safety limitsthat are different from the stated limits in NUREG-1537.

The GSTR will accept the statedlimits in NUREG-1 537 and change GSTR:SAR Section 13.1 to read, "Fuel temperature limits of 1,150 °C (if'cladding temperature is at or less than 500 °C) or 950

°C (if cladding temperature is greater than 500*°C) for stainless steel clad fuel and 500 °C for aluminum clad fuel have been setto preclude the loss of clad integrity."

14.2 Within Chapter 13, the assumed power level or trip setpoint for accident analysis is set at 1.0 MW (Maximum Hypothetical Accident, Loss of Coolant Accident) and 1.06 MW (uncontrolled rod withdrawal). However, the Limiting Safety System Setting (LSSS) and SCRAM setpoints are set at 1.1 MW. Please describe how these setpoints ensure that the safety basis is maintained.

A090

This question will be answered at a later date; however, we will change our accident analysis assumptions to be at an assumed power of 1.1 MW in our new neutronic and thermal-hydraulic analyses.

Update on neutronic/thermal hydraulic analyses-The control rod dimensions and composition and the fuel composition are finalized. Based on the work completed last month, the initial fuel enrichment in the model is set to 19.75 wt%. With this enrichment, the value of keff calculated with the control rods in the measured critical position is $0.43 above critical. Slightly adjusting the burnup of the fuel corrects for the remaining bias. Increasing the fuel burnup throughout the core by 3.5% reduced the calculated value of keff for the critical reactor to 1.00023 or $0.03 above critical with the control rods at the measured critical position. Table I shows the final predictions for the kefr values for the low-power GSTR core.

A comparison of the measured and calculated axial flux profiles is the next step in validating the model. The available experimental data for the GSTR is limited to axial flux profiles taken within the central thimble, and at an external irradiation facility not present in the MCNP model. The GSTR staff uses gold foil irradiation techniques to measure the flux as a function of the measured activation rates in gold foils. The MCNP model simulates this using a series of point detectors with appropriate multipliers at the same location as the foils in the experiment. The preliminary calculation of the flux profile in the central thimble was finished at the end of the month and is currently being evaluated and compared to the experimental data from the GSTR.

The final stage of the model validation will construct and evaluate a full power MCNP mode of the GSTR by altering the fuel, cladding, and water temperatures within the model to accurately represent the full power conditions of the reactor. Verifying the keff predictions at high power should validate that the temperatures have been correctly altered to match the operating conditions of the GSTR and that will complete the validation.

Table 1: keff a 3 Terror A from (7($) 3 Yerror Neutronic Critical ($) ($)

predictions for the low power core Rod Position Critical 1.00023 0.00007 0.00021 0.03 0.01 0.03 Rods Out 1.03650 0.00007 0.00021 5.21 0.01 0.03 Rods 0.96304 0.00007 0.00021 -5.28 0.01 0.03 Down

Sincerely, Tim DeBey USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct.

Executed on 3/28/12 Copy to:

Betty Adrian, Reactor Administrator, MS 975 USGS Reactor Operations Committee