The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVs. This specific project was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 m x 1.2 m x 17.1 cm thick [4 ft x 4 ft x 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the "mirror" insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [lo in] thick concrete wall, 2.1 m x 2.1 m [lo ft x 10 ft] square. Experiments were performed at temperature heat-upkooldown rates of 7, 14, and 2 8 " C h C12.5, 25, and SO'Fh] as measured on the heated face. A peak temperature of 454°C [850°F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code developed at Sandia National Laboratories called "Sandia One Dimensional Direct and Inverse Thermal" (SODDIT). Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing. Additionally, incident heat flux measurements can be used to design the heater system required to anneal a full-scale RPV. ~\ I Results compare favorably with those reported in NUREGKR-4212.
The objective of this project was tQ provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel could be measured during an annealing process. This project was a coordinated effort between DOE's Office of Nuclear Energy, Science and Technology; DOE's Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute's NonDestructive Evaluation Center.Ball-thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metalsheathed thermocouple probes, and air-suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as "shunting'' (electrical breakdown of insulation separating the thermocouple wires).Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454°C [850"F], all sensors measured the same temperature within about k5% (23.6"C [42.5"F]).Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball-probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball-probe temperature. The model does not predict the temperatures as well for the spring-loaded and air-suspended probes.' Retired. MASTER Investigation of RPV Temperature Measurement Methods AcknowledgmentsThis work was performed at Sandia National Laboratories, Albuquerque, New Mexico, by personnel from the Department of Energy's Light Water Reactor Technology Center, Advanced Nuclear Power Technology Department (6471) ' . . Investigation of RPV Temperature Measurement Methods Figures Executive SummaryNeutron exposure in the beltline region of a reactor pressure vessel (RPV) causes the steel to become less ductile, lowering its resistance to fiacture and making the RPV more susceptible to pressurized thermal shock (PTS). PTS occurs when an RPV is depressurized, then repressurized at a lower temperature. Annealing is the only known embrittlement management technique that restores material properties to the RPV steel. Thermal annealing returns the embrittled portions of the RPV (e.g., beltline welds] to a more ductile state and may be used to extend the service life of the plant. This investigation was undertaken to determine how accurately various methods measure the true temperature of an RPV wall during a thermal annea...
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