Tao Xu

Bio: Tao Xu is an academic researcher from Xi'an Jiaotong University. The author has contributed to research in topics: Nuclear engineering & Fault tree analysis. The author has an hindex of 1, co-authored 1 publications receiving 2 citations.

A Newly Developed Suppression Pool Model Based on the ISAA Code

Abstract: 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...

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Abstract: A simulation calculation model in the case of a severe accident such as a loss of coolant accident (LOCA) was established and calculated by MELCOR code in a marine pressurized water reactor. The effects of severe accidents on the hydrogen source term, release of the radionuclide source term, and the accident process were investigated, and the radiation consequences were analyzed. LOCA sensitivity analysis was conducted with different break sizes; the release and diffusion of source terms were analyzed. The results indicated that the hydrogen yield in the reactor core was dependent on the temperature, and on the residual water in the reactor core, which has no direct relationship with the break size. The break size directly affected the accident progression. However, when the break exceeded a certain size, it did not affect the LOCA process. The break size may not change to change the release amount and migration rule of radionuclides such as Xe and I, and the fission product CsI, as final containment is in dynamic equilibrium in LOCA conditions. The contents of fission products released into the containment vessel, major loop, cavity, and the environment are sensitive to the break size. However, no special correlation is observed between the amount released and the break size.

Severe accident analysis of the dual-functional lithium-lead test blanket module using ISAA-DFLL code

Abstract: The international thermonuclear experimental reactor (ITER) project aims to build a tokamak fusion test reactor to verify the feasibility of fusion reactors. The test blanket module (TBM) is the key component of the international thermonuclear experimental reactor. The fusion design study team proposed the concept of the dual-functional lithium-lead test blanket module (DFLL-TBM). The integral severe accident analysis (ISAA) is a self-developed ISAA code that models the progression of severe accidents in nuclear power plants. A broad spectrum of severe accident phenomena including fission product release and transport behavior is modeled in ISAA. In this paper, a new version of ISAA, referred to as ISAA-DFLL is introduced for the application of DFLL-TBM test blanket module into the treatment of multi fluids and the modules of the new physical property and heat transfer. The modification is verified by comparing the steady-state temperature distribution of the DFLL-TBM first wall with the design parameters. Then accident analysis of In-vessel loss of helium coolant in first wall and TBM pipe is conducted by using the ISAA-DFLL code. By comparing the calculation results with the general safety requirements for TBM, it is concluded that the design of the DFLL-TBM system and the modifications of the current version are reasonable and accurate.

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

Abstract: 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

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: 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.

Reliability Analysis of a Computer-Based Interlocking System with a Double 2-out-of-2 Redundancy Structure using Algebraic Binary Decision Diagrams

Abstract: The computer-based interlocking (CBI) system of urban rail is a safe-critical system. To keep the high reliability of the CBI system of urban rail, reliability analysis is very necessary and important work. The double 2-out-of-2 redun-dancy structure has been widely used in the CBI system which is a core part of control signal equipment in urban rail. The dynamic fault tree (DFT) is an extension of the static fault tree (SFT), which is widely used for the reliability modeling of dynamic systems. An algebraic binary decision diagram (ABDD) is an extension of a binary decision diagram (BDD) which introduces a kind of node denoting a sequence-dependency between events. The DFT can be evaluated by the ABDD. This paper firstly presents a reliability analysis of the CBI system in urban rail by using the ABDD-based method. The CBI system is modeled by the DFT, then the DFT is converted into a function based on a temporal algebraic framework. According to the temporal function, the ABDD can be built for reliability analysis. Compared to the Markov-based method, the ABDD-based method can avoid space-state explosion and have no limitation on the exponential distribution of component failure. A case study related to the CBI system is presented to show the advantage of using our method.