Recorded test data are presented for Test S-05-2 of the Semiscale Mod-I alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-I experiments conducted to investigate the the_rmal and hydraulic phenomena accompanying a hypothesized Joss-of-coolant' accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid t~mperature of 545°F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with. ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65°F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident. The purpose of this report is to make available the uninterpreted data from Test s-05-2 for future data analysis and test results reporting activities. The dat::t, presented in the form of graphs ih engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent.
Recorded test data are presented for Test S-06-1 of the Semtscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-1 was conducted from initial conditions of 15 568 kPa and 564 K to. . investigate the response of the Semiscale Mod-I system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injec;ted into the ,cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heat.er rods in the electrically heated core were operated at an axial peak power density which was 30% of the maximum peak power density (52.5 kW/m). The purpose of this report is to make available the uninterpreted data from Test S-06-1 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent, ..
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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