Steady-state Thermal Hydraulics Research Articles

Taillefer—A Tool for Sensitivity Analysis and Uncertainty Propagation Studies for Steady-State Thermal-Hydraulic Simulations of Involute Fuel Element Research Reactors

Taillefer is a versatile Python tool for carrying out Sensitivity Analysis (SA) and uncertainty propagation (UP) studies based on Monte Carlo sampling. Developed with the primary goal of investigating sensitivities and uncertainties of steady-state thermal-hydraulic (SSTH) safety parameters of the high-performance research reactors Forschungs Neutronenquelle Heinz Maier-Leibnitz (FRM II) in Garching, Germany, and the Réacteur à Haut Flux (RHF) in Grenoble, France, it can also be used for a large variety of other modeling problems. The work presented here aims to explain the underlying mathematical background of SA and UP studies with Taillefer and to show some steps to verify these routines. Furthermore, a real-life application example is provided that demonstrates Taillefer’s use in SSTH analysis of the RHF. For this purpose, Taillefer is coupled to the external thermal-hydraulic software PLTEMP/ANL, which is one of the codes used at FRM II and RHF to access SSTH performance and safety parameters. Determining these crucial quantities is part of identifying possible low-enriched uranium (LEU) core designs that are suitable to replace the currently used highly enriched uranium fuels of the two reactors, supporting global nonproliferation efforts. Taillefer is a powerful tool in these conversion studies, as it increases the reliability of the LEU safety parameters by providing information about sensitivities and uncertainties in addition to the nominal values predicted by the thermal-hydraulic software.

Understanding of thermal-hydraulic instabilities and the mutual interactions at supercritical pressure

In the present work, a steady state and a simple transient models are integrated to simulate various thermal-hydraulic (TH) static and dynamic instabilities in a single vertical channel, subjected to time independent power, for supercritical fluid (SCF). The scheme of U.S. supercritical water reactor (SCWR) design is analyzed for different TH instabilities and the mutual interactions of such instabilities. An in-house steady state TH model, which solves a set of nonlinear coupled mass, momentum and energy conservation equations accounting for the pressure and enthalpy dependent thermodynamic equation of state, is implemented to find the solution at the initial equilibrium condition. The steady state model is then validated against the recently available experimental results. Next, the simple transient model is integrated with the steady state model to simulate Ledinegg excursion, pressure drop oscillations (PDOs), density wave oscillations (DWOs), and the mutual interactions of each other in presence as well as in absence of a surge tank at the upstream of the heated section. A number of simulations are carried out considering various typical pump (external) characteristic curves. Results provide the understanding what are the different operational criteria on which the Ledinegg excursion, PDOs and DWOs, or the interactions of the aforementioned TH instabilities will depend.

Effect Of U-9mo/Al Fuel Densities On Neutronic And Steady State Thermal Hydraulic Parameters Of Mtr Type Research Reactor

<p>The objectives of this research work are to carry out a detailed neutronic and steady state thermal hydraulics analysis for a MTR research reactor fuelled with the low enrichment U-9Mo/Al dispersion fuels of various uranium densities. The high density uranium fuel will increase the cycle length of the reactor operation and the heat flux in the reactor core. The increasing heat flux at the fuel will causing increase the temperature of the fuel and cladding so that the coolant velocity has to be increased. However, the coolant velocity in the fuel element has a limit value due to the thermal hydraulic stability considerations in the core. Therefore, the neutronic and the steady state thermal hydraulic analysis are important in the design and operation of nuclear reactor safety. The calculations were performed using WIMS-D5 and MTRDYN codes. The WIMS-D5 code used for generating the group constants of all core materials as well as the neutronic and steady state thermal hydraulic parameters were determined by using the MTRDYN code. The calculation results showed that the excess reactivity increases as the uranium density increases since the mass of fuel in the reactor core is increased. Using the critical velocity concept, the maximum coolant velocity at fuel channel is 11.497 m/s. The maximum temperatures of the coolant, cladding and fuel meat with the uranium density of 3,66 g/cc are 70.85°C, 150.79°C and 153.24°C, respectively. The maximum temperatures are fulfilled the design limit so reactor has a safe operation at the nominal power.</p>

Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

A Large Assembly with Small Pins Concept for High-Power-Density BWRs

The large assembly with small pins (LASP) concept is an evolutionary boiling water reactor (BWR) fuel assembly design aimed at increasing the power density of BWR cores while keeping the same power-to-flow ratio, core inlet conditions, and fuel-to-moderator ratio. It is based on replacing four traditional assemblies and their large interassembly water gap regions with a single large assembly surrounded by a narrower gap region. The traditional BWR cylindrical UO2-fueled Zr-clad fuel pin design is retained, but the pins are arranged on a 22 × 22 square lattice. Twenty-five water rods within the assembly maintain the moderating power and accommodate as many finger-type control rods. The technical characteristics of LASP were evaluated and are systematically compared with a traditional 9 × 9 fuel assembly. This design study includes analyses of the steady-state thermal hydraulics, two-dimensional and three-dimensional burnup-dependent neutronics, flow-induced vibrations, and fuel pin thermomechanical behavior. Furthermore, the conceptual mechanical design of the LASP assembly is discussed. The analyses show that LASP can operate at a power density that is 20% higher than the traditional BWR assemblies while maintaining the same safety margin.

Uncertainty and sensitivity analyses for steady-state thermal–hydraulics of research reactors

The assessment of uncertainty is a desirable capability for thermal–hydraulic codes allowing the achievement of uncertainty associated with any code calculation results. In this paper a methodology, based on multivariable Taylor series expansion, suitable for introducing such a capability into a thermal–hydraulic code, method of evaluation of thermal–hydraulic parameters and the main results obtained from its application are discussed and presented. The computer code, THYC, was developed using standard correlations adopted by International Atomic Energy Agency (IAEA) for the analyses of steady-state thermal–hydraulics of research reactors. For the purpose of accuracy evaluation and showing the applicability of the proposed code, the steady-state operating conditions of the IAEA nuclear research reactor were assessed at a power level of 2 MW. The steady-state thermal–hydraulic results obtained show close agreement between the results of the present code and those of benchmark calculations. The purpose of benchmark calculations was to check how reliable the results obtained are when the code is applied to identical cases. The sensitivity study for 31 main input variables identified the most important design parameters affecting the peak clad temperature, ONB and DNB heat fluxes for plate-type research reactors. Also, the analysis used for assessing sensitivity in steady-state thermal–hydraulics of research reactors provides a means of determining which design variables are of (a) dominant, (b) significant, and (c) minor importance in severe accident decision-making.

Steady-state thermal hydraulic and safety analyses of a proposed mixed fuel (HEU&LEU) core for Pakistan research reactor-1

Pakistan Research Reactor (PARR-1) was converted from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, in 1992. The reactor is running successfully with an upgraded power level of 10 MW. In order to save money on the purchase of costly fresh LEU fuel elements, it is being thought to use some of the less burnt HEU spent fuel elements along with the present LEU fuel elements. In the present study steady-state thermal hydraulics of a proposed mixed fuel core (see Fig. 2) has been carried out. Results show that the proposed core, comprising of 24 LEU and 5 HEU standard fuel elements, with 4 LEU and one HEU control fuel elements, can be safely operated at a power level of 9.86 MW without compromising on safety. Standard computer codes and correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core and margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB).