Loss Of Coolant Accident Conditions Research Articles

Temperature increase on target due to jet impingement under BWR/PWR loss of coolant accident conditions

This report describes the temperature increase on the target plate after jet impingement on it from a ruptured pipe under BWR/PWR Loss of Coolant Accident Conditions. From test results it is shown that the temperature on the target can be conservatively estimated by taking it equal to the saturated temperature corresponding to the pressure on the target, which is given by the steam table. An experimental formula is presented to estimate the maximum temperature increase on the target.

Experimental and analytical studies of pipe whip tests under PWR LOCA conditions

Abstract A series of pipe rupture tests has been performed at the Japan Atomic Energy Research Institute (JAERI) to demonstrate the safety of primary coolant circuits in the event of pipe rupture in nuclear power plants. Pipe whip tests and jet discharge tests have been conducted under boiling water reactor (BWR) and pressurized water reactor (PWR) loss-of-coolant accident (LOCA) conditions. The present paper describes the experimental and analytical results of the pipe whip tests performed under PWR LOCA conditions using 4, 6 and 8-inch test pipes. The tests were carried out at an initial pressure and temperature of 15.7 MPa and 325°C, respectively. Moreover, a dynamic analysis of pipe whip tests was carried out using the general purpose finite element program ADINA.

Experimental study of jet discharging test results under BWR and PWR loss of coolant accident conditions

This report describes the results of the jet discharging experiments conducted at the Japan Atomic Energy Research Institute. The tests were done under BWR and PWR Loss of Coolant Accident conditions using 4 inch, 6 inch and 8 inch test pipes, and varying distance between the pipe exit and the target plate. Simple and practical experimental formulae to estimate the maximum pressure on the target plate and maximum pressure distribution are given. Further, relations between pipe reaction thrust forces and jet impingement forces are described.

Evaluation and Prediction of Critical Overhang Length Under Pipe Whip Accident

When the pipe length between break exit and restraint is long in the pipe whip accident, the pipe will undergo a plastic collapse as the moment increases. The length at which plastic collapse may occur is called the critical overhang length, (OH)cr. The experimental results of (OH)cr show good agreement with the prediction by a static simplified estimation method for (OH)cr although the pipe whipping is a dynamic phenomenon. The diagrams of (OH)cr are also described for a range of sizes of stainless steel pipe under the loss of coolant accident conditions of light water reactors.

Containment Spray Model for Predicting Radioiodine Removal in Light Water Reactors

The computer code COSMO (containment spray model) is prepared to provide information on radioiodine removal by the containment spray of boiling water reactors and pressurized water reactors involving an ice-condenser-type reactor in the event of a loss-of-coolant accident (LOCA). The code considers the removal of inorganic iodine, organic iodide, and particulate iodine in multiple compartments, taking into consideration natural deposition onto the containment wall, spray washout, liquid-film absorption, filtration, and leakage to the environment. The code is also developed by adding many abilities such as the gas convection flow between multicompartment rooms due to fluid disturbance and spray covering onto the inner structures involving the wall in the containment vessels. The calculated results are compared with the iodine removal tests of the Japan Atomic Energy Research Institute model containment, and the iodine removal for light water reactors of 1000 MW(electric) in the postulated LOCA conditions i...

Qualification of Cables to IEEE Standards 323-1974 and 383-1974

Wire and Cable manufacturers generally qualify products for class IE application by envelope type testing to user specifications and environmental conditions recommended by IEEE Standards 323-1974 and 383-1974. The General Electric Wire and Cable Business Department recently completed two such qualification programs. Cable constructions tested were 600V control cables and 600V, 2KV, and 15KV power cables insulated with flame resistant mineral filled crosslinked polyethylene. The 15KV samples included taped field splices. To determine the ability of samples to withstand a loss of coolant accident (LOCA) both early in and at the end of design life, both thermally aged and unaged cables were irradiated and subjected to a 100 day plus 10 day margin simulated LOCA test. Samples were heat aged to a time-temperature condition in excess of 40 years of service life. The aging criterion was determined from experimental data fitted to an Arrhenius curve by the Regression Method of IEEE Standard 101-72. All test specimens were irradiated using a Cobalt 60 source to a total integrated dose of 220 Mrad. The simulated LOCA conditions of one test program followed the combined requirements for Boiling Water Reactors and Pressurized Water Reactors. In the second test program, the steam pressure-temperature profile included a simulated main steam line break. Test specimens were wrapped on grounded mandrels and were electrically loaded throughout the simulated LOCA tests. After completion of environmental testing, samples were subjected to the IEEE 383 simulated post-LOCA test.

Simulated and Simultaneous Loss-of-Coolant Accident Testing of Protective Coatings for the Nuclear Industry

An apparatus was designed and constructed to test the performance of coatings under conditions of high-pressure steam and radiation that might exist under a loss-of-coolant accident (LOCA). Results from the ''simultaneous'' exposure of coatings to high-pressure steam and radiation are compared to results obtained from the conventional ''simulated'' test procedures. Coating systems were selected that had a history of performing well under simulated LOCA conditions and included zinc-based, epoxy, and phenolic primers with phenolic and modified phenolic topcoats. Coatings were exposed to $sup 60$Co radiation doses in the range of 10$sup 8$ to 10$sup 9$ rad. The study showed that the conventional simulated LOCA conditions were more severe on the coatings than those tested under the simultaneous exposure to high-pressure steam and $sup 60$Co radiation. It was concluded that coatings that satisfactorily passed the simulated LOCA tests will also pass the simultaneous LOCA tests.

An experimental study on the burst strength of fuel cladding tubes under loss of coolant accident conditions in water cooled power reactors

Experimental study has been carried out on the burst strength of fuel cladding tubes of stainless steel and zircaloy under simulated loss-of-coolant accidents in water-cooled power reactors. The results of high-temperature tensile tests and high-temperature burst tests conducted under static and dynamic loading conditions are analysed. The effect of absorbed hydrogen on the mechanical properties of the tubes is investigated.

An experimental study on the burst strength of fuel cladding tubes under loss of coolant accident conditions in water cooled power reactors☆

Experimental study has been carried out on the burst strength of fuel cladding tubes of stainless steel and zircaloy under simulated loss-of-coolant accidents in water-cooled power reactors. The results of high-temperature tensile tests and high-temperature burst tests conducted under static and dynamic loading conditions are analysed. The effect of absorbed hydrogen on the mechanical properties of the tubes is investigated.