CFD simulation of the IAEA 10 MW generic MTR reactor under loss of flow transient

Three-dimensional simulation of the IAEA 10 MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow accidents (FLOFA) and slow loss of flow accidents (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The transients were initiated from a power of 12 MW with a flow trip point at 85% nominal flow and a 200 ms time delay. The simulation shows comparable results as those published by other research groups. However, interesting 3D patterns are shown that are usually lost based on the one-dimensional simulations that other research groups have introduced. In addition, information about the maximum clad surface temperature, the maximum fuel element temperature as well as the location of hot spots in fuel channel is also reported.     

Heat transfer during boiling of a liquid metal during emergency cooldown of a fast-neutron reactor

The physics of the processes, the characteristics, and the stability of different regimes, of boiling (nucleate, projectile, disperse-ring), which are observed in experiments investigating the boiling of liquid-metal coolant in a model of a fuel assembly for a fast-neutron reactor in the emergency cooldown regime with low circulation velocity, are analyzed. The experimental setup, the, methods for performing measurements, and the experimental data on the boiling of a liquid metal are described. A mathematical model of the process of boiling of a liquid-metal, coolant in a natural-circulation loop is described, and the results of test calculations for regimes with an increase in heating and with sharp pressure drop are prresented. 7 figures, 12 references. State Science Center of the Russian Federration–A. I. Leipunskii Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 337–342, November, 1999.     

Passive modular gas safety system for a reactor

Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center, RNTs. Translated from Atomnaya Énergiya, Vol. 75, No. 1, pp. 8–13, July, 1993.     

Achievement of 1.9 MW for 300 s during the commissioning of the KSTAR ICRF transmitter

KSTAR (Korea Superconducting Tokamak Advanced Research) is a national tokamak aiming at the high beta operation based on AT (Advanced Tokamak) scenarios in Korea and ICRF (Ion Cyclotron Ranges of Frequency) is one of the essential heating and current drive tools to achieve this goal. The ICRF heating and current drive scenario requires 4 units of 2 MW transmitters with a frequency range from 25 to 60 MHz. The first KSTAR transmitter is a modified FMIT (Fusion Material Irradiation Test) transmitter consisting of four amplifier stages. An amplitude-modulated 1 mW frequency source drives a 500 W solid state wideband amplifier, which in turn drives three tuned triode/tetrode amplifier stages. The tube employed in the final power amplifier is a 4CM2500KG tetrode fabricated by CPI (Communications & Power Industries). After the fabrication of the cavity and power supply was completed in 2004, several failures of the tube during a factory and a site acceptance test occurred before eventually achieving 1.9 MW for 300 s at 33 MHz in 2007. The electrical efficiency of the FPA (Final Power Amplifier) is about 70%. Although this is a very encouraging result for the development of an ICRF transmitter for ITER (International Thermonuclear Experimental Reactor), continued efforts for a reliable operation are required to achieve the final goals of the KSTAR and ITER ICRF system.     

Numerical design of a 20 MW lead–bismuth spallation target with an injection tube

A spallation target system is a key component to be developed for an accelerator driven system (ADS). It is known that a 15–25 MW spallation target is required for a practical 1000 MW_th ADS. The design of a 20 MW spallation target is very challenging because more than 60% of the beam power is deposited as heat in a small volume of the target system. In the present work, a numerical design study was performed to obtain the optimal design parameters for a 20 MW spallation target for a 1000 MW_th ADS. A dual injection tube was proposed for a reduction of the lead–bismuth eutectic (LBE) flow rate at the target channel. The results of the present study show that a 30 cm wide proton beam with a uniform beam distribution should be adopted for a spallation target of a 20 MW power. When the dual LBE injection tube is employed, the LBE flow rate could be reduced by a factor of 7 without reducing the allowable beam current.     

5 MW CW supply system for the ITER gyrotrons Test Facility

ECH (Electron Cyclotron Heating) for ITER will deliver into the plasma 20 MW of RF power. The procurement of the RF sources will be shared equally between the three following partners: Europe, Japan and Russia. Moreover, Europe decided to develop a RF source capable of 2 MW CW of RF power, based on the design of a coaxial gyrotron with a depressed collector. In order to be able to develop and test these RF sources, a Test Facility (TF) has been built at the CRPP premises in Lausanne (CH).The present paper will first remind the main operation conditions considered to test safely a gyrotron. The power supplies parameters allowing to fulfill these conditions will be reviewed. The core of the paper content will describe the newly installed Main High Voltage Power Supply (MHVPS), to be connected to the gyrotron cathode and capable of ?60 kV/80 A-CW. The principle, the characteristics, the on-site test results will be described at the light of the requirements imposed by the gyrotron testing. Particular aspects of the installation and commissioning on-site will be highlighted in comparison with the ITER environment. The synchronized operation of the MHVPS and the BPS (Body Power Supply) on dummy load, piloted through the TF remote control, will be presented and commented.Since the TF supply structure has been built integrating the particular conditions and requirements expected for ITER, a conclusion will summarize the performances obtained at the light of these criteria.     

Thermodynamic performance of a gas core fission reactor

In this work we investigated the thermodynamic behaviour of fission products and plutonium as obtained in a gas core fission reactor with graphite walls and operated at 1200 MW thermal power. Equilibrium compositions of the system U-C-F-Pu-fission products were calculated for pressures of 0.1 MPa and 2.5 MPa and temperatures of 1300 K to 10000 K. We found that the reactor can be operated at a pressure of 2.5 MPa and a wall temperature of 2500 K without condensation of any component; no carbides are formed. The main plutonium compound is PuF₄ which, from thermodynamic point of view, can be recycled with UF₄.     

Overpressure protection analysis of an advanced integral reactor

Overpressure protection analysis of KAERI's advanced integral reactor, which has been developed to verify the performance of the System integrated Modular Advanced ReacTor (SMART), has been performed using the Transients And Setpoint Simulation/Small and Medium Reactor (TASS/SMR) code. In the analysis, the loss of feed-water and the regulating bank withdrawal events on behalf of the decrease in the heat removal by the secondary system and the reactivity and power distribution anomalies are selected as the initiating events for the analysis because the highest peak pressures of the primary system occur during these events. Conservative assumptions and the various initial/boundary conditions have been applied to the overpressure protection analysis for the advanced integral reactor. Although the pressurization of the primary system occurs due to an unbalance between the power generation in the core and the heat removal through the steam generator, the peak pressures in the cases of using the loss of feed-water and the regulating bank withdrawal event as an initiating event are well below the acceptance criteria of 18.7 MPa, due to the reactor protection system and three pilot operated safety relief valves installed in the advanced integral reactor.     

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.     

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.     

Neutronic performance of new coolants in a fusion–fission (hybrid) reactor

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li₂₀Sn₈₀ have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation.     

非能动安全系统在200MW核供热堆中的应用

200MW核供热堆主回路系统、余热排出系统和注硼系统都没有驱动设备,主回路和余热排出系统的流体流动依靠自然循环,注硼系统的注硼依靠重力。本文描述了这些系统的设计和固有安全特性。     

Nuclear steam supply system with an integral nuclear reactor for a floating NPP

The paper presents the results of conceptual design of the integral reactor plant of enhanced safety for a small-size floating NPP which is transported to the operation site in the state most prepared for operation and which is designed for power supply to remote and not easily accessible areas. Schematic flow diagrams, design and layout of NSSS and reactor plant as a whole are presented, as well as basic specifications.     

非能协安全系统在200MW核供热堆中的应用

200MW核供热堆主回路系统、余热排出系统和注硼系统都没有驱动设备,主回路和余热排出系统的体流动依靠自然循环,注硼系统的注硼依靠重力。本文描述了这些系统的设计和固有安全特性。     

Modeling aspects in linear stability analysis of a self-pressurized, natural circulation integral reactor

The stability of a self-pressurized natural circulation integral reactor is studied by means of a linear approach, taking the CAREM-25 reactor as reference.A thermohydraulic code has been improved for analysis of linear stability, great emphasis having been placed on the minimization of numerical diffusion and integration errors. A linearization method is implemented by means of numerical perturbations. The results are obtained within the frequency domain. The code is compared to a simpler analytical model, by contrasting stability maps obtained from both models for a test configuration, showing good agreement.In this type of reactor, oscillations are promoted by the two-phase regime in its long riser, and take place due to the counteraction between mass flow and buoyancy force.The stability of the system is strongly influenced by the steam-dome dynamics. Condensation in the steam zone, together with reactor power, determines the dynamical state of the system.The phase-lag introduced by the core dynamic regarding the riser timing, together with the sensitivity of the buoyancy force due to flow changes, determines the sustainability of the oscillation. A parametric study is carried out, gradually increasing the complexity of the model, to analyze the influence of different factors on the oscillation sustainability, concerning physical process and modeling approaches. The analysis includes the relative velocities between phases, the axial power profile along the core, the buoyancy force due to subcooled density changes, the flashing effect, the core dynamic and the pressure feedback due to self-pressurization. The steam-dome-pressure feedback is identified as a stabilizing effect, as long as it decreases the sensitivity of the buoyancy force.     

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.     

Deformation analysis of a 900 MW reactor pressure vessel head by means of holographic interferometry

During pressure build-up in a 900 MW reactor pressure vessel, the head of the vessel was holographed. It will be shown how a maximum of information can be extracted from the hologram using computer generated interferograms. Based on a trial and error method the deformation assumption for the head is altered until a best correlation is reached between computer generation and experiment.     

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.