Subcritical multiplication factor and source efficiency in accelerator-driven system

The subcritical multiplication factor k_s   and the external neutron source efficiency φ^∗${\phi }^{\ast }$ are important parameters in the accelerator-driven system (ADS) performance assessment. The theoretical relation between k_s and the effective multiplication factor k_eff in a subcritical system is discussed in different cases of subcritical system. On the basis of the theoretical background, the dependence of k_s   and φ^∗${\phi }^{\ast }$ on subcriticality, source position, and energy is numerically investigated using a simple thermal subcritical model. For the sake of experimental evaluation of k_s   and φ^∗${\phi }^{\ast }$, the ADS experiments have been carried out in the subcritical systems combined with 14 MeV pulsed neutrons of the Kyoto University Critical Assembly (KUCA). The k_s   and φ^∗${\phi }^{\ast }$ parameters are successfully measured by utilizing the reaction rate distribution obtained by the optical fiber detectors in the subcritical system, within a relative difference of less than 7% and 12% for k_s   and φ^∗${\phi }^{\ast }$, respectively, between measured and calculated values for most studied cases.     

Database development for analysis of accelerator-driven systems

Nuclear data libraries for study of transmutation, activation and neutron transport in materials of accelerator-driven systems are advocated. The libraries include nuclear data describing processes in materials irradiated by intermediate and highenergy particles and cover wide range of target nuclei and incident particle energies.     

Transmutation of nuclear waste in accelerator-driven systems: Fast spectrum

The elimination of nuclear waste is currently a pressing issue in view of the global energy crisis and increasing radiotoxic waste inventories. One potential and well-researched solution to these issues is to transmute nuclear waste in a thorium-based accelerator-driven system, such as the energy amplifier.     

Transmutation of nuclear waste in accelerator-driven systems: Thermal spectrum

The TRIGA accelerator-driven experiment is a milestone in the roadmap towards the demonstration of the transmutation of nuclear waste in accelerator-driven systems (ADSs). It is designed in such a way as to couple the different components of an ADS in a realistic environment by using an existing critical reactor adapted to the experiment’s objectives, at significant power levels.     

CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.     

Comparison of the transient behaviour of LBE-and gas-cooled experimental accelerator-driven systems

The paper presents a comparison of transient calculations for two 80-MWt MOX-fuelled experimental accelerator-driven systems (XADS), one cooled by lead-bismuth eutectic and the other by helium. The results for protected (with accelerator beam trip) and unprotected (without accelerator beam trip) transient overpower, spurious beam trip, loss of flow, loss of heat sink, and loss of coolant accidents, as well as the failure of heat exchanger tubes, are analysed for the two systems. The analysis was performed using TRAC/AAA, which is part of the PSI FAST code system. The advantages and shortcomings of the two designs from the viewpoint of their transient behaviour are discussed.     

Mathematical modeling of compression wave formation in the interaction of melted fuel and coolant

Moscow Engineering Physics Institute. Nuclear Reactor Institute of the "Kurchatov Institute" Scientific Center. Translated from Atomnaya Énergiya, Vol. 75, No. 3, pp. 175-179, September, 1993.     

Corrosion mitigation in coolant systems in nuclear power plants

Nuclear power plants have suffered various failures through corrosion causing economic losses, increasing the radiation exposure to personnel and increasing the possibility of environmental risk. Many examples of different corrosion mechanisms and their consequences for nuclear power plant (NPP) working conditions are recognized and described. Nevertheless, several issues related to the corrosion of materials used for NPP constructions are still unexplained. This paper gives short, basic information about selected methods of the corrosion reduction and corrosion inhibitors used in coolant systems in nuclear power plants, mainly in pressurized water reactors PWRs and boiling water reactors BWRs. Present data are based in the open scientific and technical literature since 1990.     

Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% ²³⁵U enriched UAl_x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% ²³⁵U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE.Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.     

Probabilistic evaluation of core-internal coolant flow distribution with correlation among fuel subassemblies

The effect of local and global uncertainties in properties and tolerances on the probabilistic variation in coolant flow rate of the hottest channel in a prototype FBR core is discussed in terms of the coolant flow correlation among all the core-internal coolant channels. Sample calculations of the probabilistic deviations in subassembly flow rate for a prototype FBR were about 30% smaller than those conventionally calculated.     

Chemistry control analysis of lead alloys systems to be used as nuclear coolant or spallation target

This study presents the lead alloy system chemistry analysis for use as nuclear coolant or spallation target in ADS related systems in order to set down the needs for purification processes and monitoring. The study is limited here to the two main impurities, oxygen and iron. The analysis of the various potential pollution sources that may occur during the various operating modes is given, as well as a first pollution rate assessment. In order to limit the consequences in term of contamination (clogging) and corrosion, it is necessary to define specifications for operation as regards oxygen and iron content in the fluid. As iron cannot be measured and controlled up to now, the best specification is to set the oxygen as high as possible, defined by the cold leg interface temperature to ensure tolerable contamination, in order to maximize the oxidation area to ensure corrosion protection by self-healing oxide layer for the entire system.     

Boiling water reactor fuel heatup during a loss of coolant while refueling

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.     

Consequences of a step variation in coolant temperature on a fuel rod with cladding

Time dependent temperature distributions in cylindrical fuel rods with cladding are evaluated analytically. The transient condition is due to a step variation in coolant temperature. Some numerical results are related and a short discussion is introduced.     

Determination of the region of vibroacoustic resonances of coolant and fuel assemblies in advanced enhanced-power reactors

Computational methods and algorithms for determining the region of vibroacoustic resonances of fuel elements and fuel assemblies for existing and planned VVER, VVER-SKD have been developed. The results obtained are used to analyze the region of vibroacoustic resonances of the coolant and fuel assemblies in the new-generation reactors VVER-1200, -1700.     

Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the subassemblies with high precision.

In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift.

The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow.

Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power.     

Transient gasification in an NBG-18 coolant channel of a VHTR prismatic fuel element

This paper investigates the transient gasification of NBG-18 nuclear graphite with atmospheric air ingress in a 0.8-m long coolant channel of a prismatic Very High Temperature Reactor fuel element. Analysis varied the initial graphite and air inlet temperature, T_o, from 800 to 1100 K at air inlet Reynolds number, Re_in = 5, 10 and 20. The analysis employs a Generic Interface that couples a multi-species diffusion and flow model to readout tables of the CO and CO₂ production fluxes. These fluxes are functions of the graphite local surface temperature, oxygen partial pressure and graphite weight loss fraction and calculated using a chemical-reactions kinetics model for the gasification of nuclear graphite. The analysis accounts for the heats of formation of CO and CO₂ gases, the heat conduction in the graphite sleeve, and the change in the oxygen partial pressure in the bulk gas flow mixture along the channel. Transient calculations performed up to a weight loss fraction of 0.10 at the entrance of the channel, t₁₀. They include the local graphite surface temperature and composition of bulk gas flow, the local and total graphite weight losses and the local and total production rates of CO and CO₂ gases. The heat released in the exothermic production reactions of these gases increases the local graphite surface temperature, accelerating its gasification. At the end of the calculated gasification transient, t = t₁₀, the graphite weight loss is highest at the channel entrance and decreases rapidly with axial distance into the channel, to its lowest value where oxygen in the bulk gas flow is depleted. Increasing T_o decreases t₁₀ and the total graphite loss, while increasing Re_in decreases t₁₀ but increases graphite loss.