This study describes the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to Station Blackout, defined as a loss of offsite power combined with failure of all onsite emergency diesel-generators to start and load. Every effort has been made to employ the most realistic assumptions during the process of defining the sequence of events for this hypothetical accident. DC power is assumed to remain available from the unit batteries during the initial phase and the operator actions and corresponding events during this period are described using results provided by an analysis code developed specifically for this purpose. The Station Blackout is assumed to persist beyond the point of battery exhaustion and the events during this second phase of the accident in which dc power would be unavailable were determined through use of the MARCH code. Without dc power, cooling water could no longer be injected into the reactor vessel and the events of the second phase include core meltdown and subsequent containment failure. An estimate of the magnitude and timing of the concomitant release of the noble gas, cesium, and iodine-based fission products to the environment is provided in Volume 2 of this report. 58 refs., 75more » figs., 8 tabs.« less

Station Blackout at Browns Ferry Unit One - accident sequence analysis. Volume 1

Performance Evaluation of Accident Tolerant Fuel under Station Blackout Accident in PWR Nuclear Power Plant by Improved ISAA code

Study on the effect of flow blockage due to rod deformation in QUENCH experiment

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. Flashing from coolant release, condensation heat transfer, intercompartment transport, and engineered safety features are described using best estimate models and correlations often based upon experiment analyses. Two notable capabilities of PACER that differ from most other containment loads codes are the modeling of the rates of steam and water formation accompanying coolant release as well as the correlations for steam condensation upon structure.

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PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 1: Code models and correlations

Abstract For severe accidents, in-vessel retention (IVR) is a very effective and crucial severe accident mitigation measure. The lower head of the reactor pressure vessel plays a vital role in the IVR strategy. The failure of the lower head may lead to the release of radioactive substances into the environment. During the implementation of IVR, the lower head is in a high-temperature environment, and its main failure form is creep failure. Therefore, to ensure the successful implementation of the IVR strategy and prevent radioactive material leakage, it is necessary to conduct an in-depth analysis of the lower head. In this paper, the lower head thermal-mechanical creep failure (LHTCF) module is developed based on the theory of plate and shell and Norton-type constructive creep laws. Through the mechanical analysis of the lower head, seven failure criteria are used to evaluate the integrity of the lower head. Finally, the LHTCF module is integrated into the integrated severe accident analysis (ISAA) program, and the accuracy of the module is validated by numerical calculation of the Organisation for Economic Co-operation and Development Lower Head Failure (OLHF) experiment. Through the comprehensive judgment of different failure criteria, the final simulation results are in good agreement with the experimental data. The results show that the wall thickness at the crack decreases sharply before failure due to the effect of creep, and the stress increases abruptly at the failure time. The LHTCF module developed in this paper can accurately predict the creep behavior of the lower head, and the calculated failure time, position, and thickness distribution agree well with the experimental results.

Development and Validation of Thermal-Mechanical Creep Failure Module for Reactor Pressure Vessel Lower Head

The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas fl...

Jian Deng

Papers

A Newly Developed Suppression Pool Model Based on the ISAA Code