The coronal-radicular amputation or radicular hemisection is defined as the sectioning fragments coronal-radicular of the lower molar with clinical damage followed endodontic treatment and prosthetics rehabilitation. This clinical treatment is viable in presence of the radicular decay or furca damage. This is a report case of radicular hemisection of lower molar with decay and bone loss that compromise distal root. The objective was elimination of distal root and conserved mesial root with endodontic and prosthetics treatment.
were performed as part of the Semiscale Mod-1 portion of the Semiscale Program conducted by EG&G Idaho ,Inc., for the United States Government. These tests are part of the steam generator tube rupture test series (Test Series 28) performed to investigate the response ,of the Mod-1 system to steam generator tube ruptures during a hypothesized loss-of-coolant accident (LOCA). The specific objective of these tests was to refine the definition of the lower limit of steam generator tube ruptures at which high peak cladding temperatures occur, as set by Test S-28-2. Hardware configuration and test parameters were selected to yield a system response that simulates the response of a pressurized water reactor during a hypothesized LOCA with subsequent refill and reflood.These tests utilized the Semiscale Mod-1 system equipped with a pressure vessel with a 40-rod electrically heated core; an intact loop with pump, steam generator, and pressurizer; a broken loop with simulated pump, simulated steam generator, and rupture assemblies; and a pressure suppression system with header, pressure suppression tank, and heated steam supply system. High and low pressure coolant injection pumps and a coolant injection accumulator were provided for each system loop. An additional injection accumulator was p;ovided for the intact loop hot leg. The intact loop hot leg injection flow rate was set to simulate the rupture of 16 (Test S-28-8), 12 (Test S-28-10), and 14 (Test S-28-11) steam generator tubes. For the tests, four heater rods were intentionally unpowered to simulate the effect of control rod guide tubes and the power in three heater rods was increased to produce a slightly peaked power profile.
Rccorded test data are presenred Tor Tesls 3-28-7, 5-28-9, and S 23 12 of the Semiscale Mod-l steam generator Lube ~u y t u r e test scric~. Tllece t e~t~ arc among sever21 Semiscafe Mod-1 expenments conducted Io investigate thc thcrmnl and hydrai~lic. phenomena accompanying a hypothesized loss-of-coolant acciderit in a pressurizcd water reactor (PWR) system.Tests S-28-7, S-28-9, and S-28-12 were conducred l k o r~~ illilia1 ~u~~d i t i o~l s of 1.5 736 kPa and 557 K, 1 5 754 kPa and 556 K, and-15 704 kPa and 559'K, respectivcly, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop coldbleg-piping.The specific-objective of these tests was to refine the definition of the upper limit-of steam generator tube ruptures at which high peak cladding temperaturgs occur, as set by Test S-28-1. During rhese tests, cuvli~lg water was injccted into the cold leg of the inlact and broken loops to simulate emergency core coolant in a PWR. Thirty (Test S-28-7), 34 (Test S-28-9), and 20 (Test S-28-12) steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.The purpose of this report is to make available the u~i i n t e r p~t t t d data f~o m Tests S-28-7, S-28-9, and S-28-12 for future data analysis and test reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only t o the extent necessary to ensure that they are reasonable and cul~sistent. SUMMARYTests S-28-7, S-28-9, and S-28-12 were Performed as part of the ~emisca1e"~od-1 portion of the Semiscale Program conducted by EG&G Idaho, Inc., for the United States Government. These tests are part of the steam generator tube rupture test series (Test Series 28) performed to investigate the response of the Mod-1 system to steam generator tube ruptures during a hypothesized loss-of-coolant accident (LOCA). The test objective specific to Tests S-28-7, S-28-9, and S-28-12 was to refine the definition of the upper limit of steam generator tube ruptures at which high peak cla'dding temperatures occur, as set by Test S-28-1. Hardware configuration and test parameters were selected to yield a system response that simulates the response of a pressurized water reactor during a hypothesized LOCA with subsequent refill and reflood.These tests utilized the Semiscale Mod-1 system equipped with a pressure vessel with a '40-rod electrically heated core; an intact loop with pump, steam generator, and pressurizer; a broken loop with simulated pump, simulated steam generator, and rupture assemblies; and a pressure suppression system with header, pressure suppression tank, and heated steam supply system. High and low pressure coolant injection pumps and a coolant injection accumulator were provided for each system loop. An additional injection accumulator.was provided for the intact loop hot leg. The intact loop hot leg injection flow rate was set to simulate the rupture ...
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