Separate-effect experiment simulating steam direct-contact condensation on ECCS （emergency core cooling system） water in PWR （pressurized water reactor） cold legs during reflood phase of large-break LOCA （loss-of...Separate-effect experiment simulating steam direct-contact condensation on ECCS （emergency core cooling system） water in PWR （pressurized water reactor） cold legs during reflood phase of large-break LOCA （loss-of-coolant accident） was conducted in OECD/NEA ROSA Project using the LSTF （large scale test facility）. A new test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Clear images of high-speed video camera were successfully obtained on droplet behaviors through the viewer of the test section, especially for annular mist flow.展开更多
The QUENCH experimental programme at Karlsruhe under severe accident conditions, but while the geometry is still Institute of Technology investigates heat-up and reflooding of a core mainly rod-like. The recent QUENCH...The QUENCH experimental programme at Karlsruhe under severe accident conditions, but while the geometry is still Institute of Technology investigates heat-up and reflooding of a core mainly rod-like. The recent QUENCH-ACM series of experiments, comprising QUENCH-12 （El 10 cladding alloy）, -14 （M5 alloy） and -15 （Zirlo^TM alloy）, together with QUENCH-06 （reference case, Zircaloy-4 alloy） addressed the effect of alternative cladding materials on oxidation and quenching under similar conditions. Superficial inspection of the experimental results reveals only minor differences in the thermal and oxidation response, except for the much larger hydrogen release during reflood in QUENCH-12. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2, modified to represent the QUENCH facility and to invoke alternative oxidation correlations. The calculations agreed rather well with experiments QUENCH-06, -14 and -15, but the significant hydrogen release during reflood in QUENCH-12 was not captured. Closer examination of the experimental results reveals further differences between QUENCH-12 which may be linked to breakaway oxidation of the E110 cladding. The analyses support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5 or ZirloTM, but the E-110 exhibited a contrasting behaviour with a consequent impact on the reflooding.展开更多
In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead ...In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a ＂debris bed＂. The typical particle size in a debris bed might reach few millimeters （characteristic length-scale： 1-5 mm）. The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN （Institut de radioprotection et de surete nucleaire）, to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow （water/steam）, in a porous medium （diameter 180 mm, height 200 mm）, initially heated to a high temperature （400℃ or 700℃）. On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.展开更多
文摘Separate-effect experiment simulating steam direct-contact condensation on ECCS （emergency core cooling system） water in PWR （pressurized water reactor） cold legs during reflood phase of large-break LOCA （loss-of-coolant accident） was conducted in OECD/NEA ROSA Project using the LSTF （large scale test facility）. A new test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Clear images of high-speed video camera were successfully obtained on droplet behaviors through the viewer of the test section, especially for annular mist flow.
文摘The QUENCH experimental programme at Karlsruhe under severe accident conditions, but while the geometry is still Institute of Technology investigates heat-up and reflooding of a core mainly rod-like. The recent QUENCH-ACM series of experiments, comprising QUENCH-12 （El 10 cladding alloy）, -14 （M5 alloy） and -15 （Zirlo^TM alloy）, together with QUENCH-06 （reference case, Zircaloy-4 alloy） addressed the effect of alternative cladding materials on oxidation and quenching under similar conditions. Superficial inspection of the experimental results reveals only minor differences in the thermal and oxidation response, except for the much larger hydrogen release during reflood in QUENCH-12. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2, modified to represent the QUENCH facility and to invoke alternative oxidation correlations. The calculations agreed rather well with experiments QUENCH-06, -14 and -15, but the significant hydrogen release during reflood in QUENCH-12 was not captured. Closer examination of the experimental results reveals further differences between QUENCH-12 which may be linked to breakaway oxidation of the E110 cladding. The analyses support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5 or ZirloTM, but the E-110 exhibited a contrasting behaviour with a consequent impact on the reflooding.
文摘In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a ＂debris bed＂. The typical particle size in a debris bed might reach few millimeters （characteristic length-scale： 1-5 mm）. The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN （Institut de radioprotection et de surete nucleaire）, to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow （water/steam）, in a porous medium （diameter 180 mm, height 200 mm）, initially heated to a high temperature （400℃ or 700℃）. On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.
