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A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility of ROSA-V Program to have an insight into effects of accident management action on core cooling during a simulated vessel top break loss-of-coolant accident with a total failure assumption on the high pressure injection (HPI) system at a pressurized water reactor (PWR). Typical phenomena of vessel top break with break sizes between 1.0 and 0.1% cold leg break equivalent were clarified including upper head water level transients related to steam discharge, coolant mass inventory related to core heat-up, performance of core exit thermocouple (CET)and three-dimensional steam flows in core and core exit. Both operator actions of HPI recovery in the 1.0% top break and steam generator depressurization in the 0.1% top break resulted in immediate recovery of core cooling when these were initiated by CET heat-up at 623 K., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2009-057

This report summarizes performances of core exit thermocouples (CETs) observed in 12 ROSA/LSTF tests which include ten small-break loss-of-coolant accident (SBLOCA) tests and two abnormal transient tests as an additional report to the OECD/NEA ROSA Project Test 6-1 report. The contents of this report are prepared to a task group which was set up to review and consolidate background knowledge of CET application to PWR accident management (AM) measures in April 2008 in the Working Group of Analysis and Management of Accident (WGAMA) at OECD/NEA. These tests cover wide ranges of test conditions such as size and location of break, primary pressure, core power, reflux water fall-back and operator actions. CET performances relative to the core temperature history are studied in each test, and general performances of CET are summarized focusing on the time delay and slow and low temperature excursion., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2009-011

This report presents major results observed in LOCA test (SB-CL-09) conducted at the ROSA/LSTF test facility simulating 10% cold leg break in a 4-loop Westinghouse-type PWR. Following are found in this test with an assumption of high pressure injection system. (1) The relatively large break size resulted in pressure inverse within 2 minutes between the primary and steam generator secondary sides. (2) During a loop-seal clearing (LSC) process started at about 1 minutes after the break, the core water level was suppressed to almost lower end and then it recovered to the middle core height. The water level remained at the middle height was due to remained water levels in the SG U-tube inlet sides which were higher than their outlet sides. (3) Significant core heat-up was observed above the water level at the middle height and core power was tripped off at 111s. (4) The effects of fall-back water from the intact loop hot leg was observed by the local core cooling., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2008-087

A small break LOCA experiment (SB-PV-06) was conducted at the LSTF of ROSA-V program to study effects of rapid secondary depressuriza-tion action on core cooling as one of accident management (AM) measures for a PWR in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. The break simulated 10 instrument tubes rupture equivalent to 0.2% cold leg break. It was clarified through comparison with former experiments that (1) the depressurization initiated by detecting the vessel level below the primary loop (4545s) was degraded by the gas inflow resulting in whole core uncovery prior to the start of low pressure injection and (2) an alternative start of the depressurization by detecting level decrease at the SG outlet plenum (2330s), would limit the core uncovery suggesting more effective parameter for the AM measures. The report presents the experiment results with the effects of rapid depressurization initiation timing., A small break LOCA experiment (SB-PV-06) was conducted at the LSTF of ROSA-V program to study effects of rapid secondary depressuriza-tion action on core cooling as one of accident management (AM) measures for a PWR in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. The break simulated 10 instrument tubes rupture equivalent to 0.2% cold leg break. It was clarified through comparison with former experiments that (1) the depressurization initiated by detecting the vessel level below the primary loop (4545s) was degraded by the gas inflow resulting in whole core uncovery prior to the start of low pressure injection and (2) an alternative start of the depressurization by detecting level decrease at the SG outlet plenum (2330s), would limit the core uncovery suggesting more effective parameter for the AM measures. The report presents the experiment results with the effects of rapid depressurization initiation timing., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2007-037

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup., A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2006-018

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system., A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system., 著者所属: 日本原子力研究開発機構(JAEA), JAEA-Research 2006-072