Thermal-hydraulic modeling for deterministic safety analysis of portable equipment application in the VVER-1000 nuclear reactor during loss of ultimate heat sink accident using RELAP5/MOD3.2 code

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

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Thermal Hydraulic Analysis of Severe Accident in PFBR

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B ◽

10.1115/icone18-29356 ◽

2010 ◽

Author(s):

K. Velusamy ◽

P. Chellapandi ◽

G. R. Raviprasan ◽

P. Selvaraj ◽

S. C. Chetal

Keyword(s):

Atmospheric Pressure ◽

Fluid Dynamic ◽

Hydraulic Modeling ◽

Severe Accident ◽

Maximum Temperature ◽

Maximum Pressure ◽

Dynamic Calculation ◽

Thermal Hydraulic Modeling ◽

Thermal Hydraulic Analysis ◽

Design Pressure

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

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Thermal-hydraulic modeling and analysis of spool valve with sloping U-shape notch by bond graph

Journal of Central South University ◽

10.1007/s11771-015-2968-x ◽

2015 ◽

Vol 22 (11) ◽

pp. 4205-4212

Author(s):

Lei Lou ◽

Wan-rong Wu ◽

Zhao-Qiang Wang ◽

Xiang-jing Liang

Keyword(s):

Hydraulic Modeling ◽

Bond Graph ◽

Modeling And Analysis ◽

Thermal Hydraulic Modeling ◽

Spool Valve

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Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp190108029a ◽

2019 ◽

Vol 34 (3) ◽

pp. 238-242

Author(s):

Rex Abrefah ◽

Prince Atsu ◽

Robert Sogbadji

Keyword(s):

High Pressure ◽

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Nuclear Reactor ◽

Atomic Energy Commission ◽

Clean Energy ◽

Safety Analysis ◽

Operating Conditions ◽

The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

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Thorium as a nuclear material: physics features and a safety analysis due to insertion of ThO2 in a PWR nuclear reactor fuel

Brazilian Journal of Radiation Sciences ◽

10.15392/bjrs.v8i3.1178 ◽

2020 ◽

Vol 8 (3) ◽

Author(s):

Clóves Júnior Da Fonseca ◽

Cláudio Luiz De Oliveira ◽

Marcos Paulo Cavaliere De Medeiros ◽

Eduardo Henrique Fernandes Fonseca ◽

Camila Oliveira Baptista

Keyword(s):

Nuclear Reactor ◽

Safety Analysis ◽

Nuclear Material ◽

Reactor Fuel ◽

Nuclear Reactor Fuel ◽

Material Physics

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Experiments on Gas Mixing and Stratification Driven by Jets and Plumes in Large-Scale, Multi-Compartment Geometries

Volume 2: Thermal Hydraulics ◽

10.1115/icone14-89290 ◽

2006 ◽

Author(s):

Robert Zboray ◽

Domenico Paladino ◽

Olivier Auban

Keyword(s):

High Resolution ◽

Time Evolution ◽

Large Scale ◽

Nuclear Reactor ◽

Safety Analysis ◽

Mixing Process ◽

Gas Mixing ◽

Computational Tools ◽

Different Gases

The present paper discusses experiments carried out to examine mixing of different gases (steam, air) and the evolution their distributions in large-scale, multi compartment geometry imitating nuclear reactor containment compartments. The flow and the mixing process in the experiments are driven by plumes and jets representing source structures with different momentum-to-buoyancy strength. The time evolution of the relevant parameters like gas concentrations, velocities and temperatures are followed using dedicated instrumentation. The data obtained is meant to be used for the validation and development of high-resolution, mainly CFD based, 3D computational tools for nuclear reactor containment safety analysis.

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