Sensitivity Analysis and Uncertainty Propagation from Basic Nuclear Data to Reactor Physics and Safety Relevant Parameters
نویسنده
چکیده
Reliable knowledge of the uncertainties in important reactor parameters, like criticality, radiation load on the reactor components, neutron/gamma ray flux, nuclear heating and dose are crucial for nuclear safety concerns. Uncertainties are introduced either through the calculational algorithms or through the data uncertainties. The paper addresses the uncertainties in the reactor parameters linked to the basic nuclear data uncertainties. The method used is based on linear perturbation theory to calculate the sensitivity coefficients, and propagates these sensitivities using the basic data covariance matrices to the target reactor quantities.
منابع مشابه
Overcoming the uncertainty in a research reactor LOCA in level-1 PSA; Fuzzy based fault-tree/event-tree analysis
Probabilistic safety assessment (PSA) which plays a crucial role in risk evaluation is a quantitative approach intended to demonstrate how a nuclear reactor meets the safety margins as part of the licensing process. Despite PSA merits, some shortcomings associated with the final results exist. Conventional PSA uses crisp values to represent the failure probabilities of basic events. This causes...
متن کاملThermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model
Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very f...
متن کاملThe BEMUSE Programme : Best - Estimate Methods Uncertainty and Sensitivity Evaluation - Phase
The BEMUSE (Best Estimate Methods – Uncertainty and Sensitivity Evaluation) Programme has been promoted by the Working Group on Accident Management and Analysis (GAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI) [1]. The high-level objectives of the work are: • To evaluate the practicability, the quality and the reliability of Best-Estimate (BE) methods including...
متن کاملNuclear Fuel Safety Threshold Determined by Logistic Regression Plus Uncertainty
Abstract—Analysis of the uncertainty quantification related to nuclear safety margins applied to the nuclear reactor is an important concept to prevent future radioactive accidents. The nuclear fuel performance code may involve the tolerance level determined by traditional deterministic models producing acceptable results at burn cycles under 62 GWd/MTU. The behavior of nuclear fuel can simulat...
متن کاملUncertainty Analyses Using the MELCOR Severe Accident Analysis Code
The MELCOR code is a detailed system level computer code capable of performing integrated selfconsistent analyses of severe accident progression in commercial nuclear power plants, supporting level 2 probablilistic risk assessment (PRA) studies. Originally developed as a fast running tool with simplified models for performing probabilistic safety analyses and sensitivity studies, MELCOR now emp...
متن کامل