Joy L. Rempe

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

Initial Back-to-Back Fission Chamber Testing in ATRC

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to provide calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.

Long Duration Testing of Type C Thermocouples at 1500 °C

Experience with Type C thermocouples operating for long periods in the 1400 to 1600 °C temperature range indicate that significant decalibration occurs, often leading to expensive downtime and material waste. As part of an effort to understand the mechanisms causing drift in these thermocouples, the Idaho National Laboratory conducted a long duration test at 1500 °C containing eight Type C thermocouples. As report in this document, results from this long duration test were adversely affected due to oxygen ingress. Nevertheless, results provide key insights about the impact of precipitate formation on thermoelectric response. Post-test examinations indicate that thermocouple signal was not adversely impacted by the precipitates detected after 1,000 hours of heating at 1,500 °C and suggest that the signal would not have been adversely impacted by these precipitates for longer durations (if oxygen ingress had not occurred in this test).

ICONE15-10842 INITIAL RESULTS FROM INVESTIGATIONS TO ENHANCE THE PERFORMANCE OF HIGH TEMPERATURE IRRADIATION-RESISTANT THERMOCOUPLES

Several options have been identified that could further enhance the lifetime and reliability of thermocouples developed by the Idaho National Laboratory (INL) for in-pile testing, allowing their use in higher temperatures applications (up to at least 1700 ̊C). A joint project between the INL and the University of Idaho (UI) is underway to investigate these options and, ultimately, provide recommendations for an enhanced thermocouple design. This paper presents preliminary results from this UI/INL effort. Tests show that unalloyed, but doped, molybdenum wires (ODS-Momolybdenum doped with lanthanum oxide and KW-Mo – molybdenum doped with silicon, tungsten and potassium) better retain ductility at higher temperatures than evaluated candidate undoped developmental alloys (Mo-1.6%Nb and Mo-3%Nb). Thermocouples that contain unalloyed molybdenum were also observed to have better high temperature resolution. Candidate niobium alloys (Nb-1%Zr, Nb-4%Mo, Nb-6%Mo, and Nb-8%Mo) became brittle at lower heating temperatures and shorter durations than any of the wires primarily containing molybdenum. Hence, results indicate that a combination of either ODS-Mo or KW-Mo with Nb-1%Zr appear to be the most favorable configuration. Initial results also show that thermocouple stability can also be improved by using larger diameter wires.

Film Boiling on Downward Quenching Hemisphere of Varying Sizes

Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Laminar Transition Apparatus). Two test sections are made of copper to maintain low Biot numbers. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of all the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the test sections. The measured data are compared with the numerical predictions from laminar film boiling analysis. The measured heat transfer coefficients are found to be greater than those predicted by the conventional laminar flow theory on account of the interfacial wavy motion incurred by the Helmholtz instability. Incorporation of the wavy motion model considerably improves the agreement between the experimental and numerical results in terms of heat transfer coefficient. In addition, the interfacial wavy motion and the quenching process are visualized through a digital camera.