• Energy Research
• 7. Clean energy close
• 12. Responsible consumption close
• Annals of Nuclear Energy close

Abstract Heat pipes and two-phase thermosyphons are highly efficient heat transfer devices utilizing continuous evaporation and condensation of working fluid for two-phase heat transport in closed systems. Because of the nearly isothermal and fully passive phase-change heat transfer mechanism, heat pipes and thermosyphons have found many applications in nuclear engineering, space technologies, and other energy systems. High-temperature heat pipes are used in nuclear microreactors to remove fission power from the primary system and are coupled with power conversion systems or process heat applications. Modeling of the two-phase flow phenomena inside a heat pipe is essential to its design and safety analysis. In this study, a comprehensive one-dimensional two-phase three-field flow model has been developed for the analysis of heat pipes in normal operation conditions and transients. The conservation or field equations of mass, momentum, and energy were developed for the liquid film, vapor, and droplet. In addition, constitutive models or correlations were reviewed thoroughly and provided for the closure of the three-field equations. Specific constitutive equations regarding interfacial mass and heat transfer at two interfaces, namely film-gas interface and gas-droplet interface, were reviewed for droplet entrainment and deposition rates as well as film and droplet evaporation rates. Additionally, mechanistic correlations of annular flow film thickness were recommended for the modeling of the thermosyphons without a wick as a critical constitutive correlation. Furthermore, experimental data needs from new experiments using a prototype working fluid or surrogate fluids for the model validation of high-temperature heat pipes in microreactors were recommended for future research.

Abstract Flow-accelerated corrosion (FAC) is a degradation mechanism that affects carbon steel piping in power plants. The failures and degradation due to FAC have necessitated numerous replacements in many power plants. Several computer codes around the world were developed as part of a systematic program or process to control FAC in power plant utilities. The typical plant model requires the input of the flow parameters, piping configuration and the plant water chemistry. The results on FAC rate are considered the key to proper selection of components for inspection. The lack of information on the effect of the upstream components located in the proximity limited the accuracy of the FAC prediction tools and hence will affect the accuracy in identifying potential inspection locations. In the present study 211 inspection data for 90° carbon steel elbows from several nuclear power plants were used to determine the effect of the proximity between two components on the FAC wear rate. The effect of the velocity as well as the distance between the elbows and the upstream components is discussed in the present analysis. Based on the analyzed trends obtained from the inspection data, significant increase in the wear rate of approximately 70% on average is identified to be due to the proximity.

Abstract The Seed-and-Blanket (SB it is few percent higher than that of the 2-stage PWR-ABR system. Approximately 7% of the thorium fed to the blanket is converted into energy, which makes the thorium fuel utilization approximately 12 times the utilization of natural uranium in PWRs. A comprehensive fuel cycle evaluation performed with the methodology developed by the recent U.S Department of Energy’s Nuclear Fuel Cycle Evaluation and Screening campaign concludes that the PWR-S&B system has similar fuel cycle performance characteristics as the PWR-ABR system. The S&B concept may potentially feature improved economics and resource utilization relative to the ABR.

Abstract Space–time kinetics calculations for CANDU ® 1 reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the “best” effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the “best” effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the correct effective delayed-neutron fraction. It is concluded that lattice-level weighting with a detailed fine-group heterogeneous adjoint is desirable for the correct calculation of the effective delayed-neutron fraction in CANDU lattices.

Abstract With the aging of current isotope-producing reactors and the rising demand for medical isotopes, future shortages are predicted to greatly increase the cost of Mo-99. Molten salt reactors have been proposed for the production of Mo-99, and have the potential for safe, efficient operation at a variety of scales. However, while Mo-99 is a major fission product, it is unclear how accessible it will be for direct extraction from a fuel salt given its relatively short half-life. In this study, a computationally efficient mathematical model was used to analyze fission product accumulation in an MSR, with the specific goal of assessing Mo-99 production by continuous or near-continuous extraction from the reactor. The model includes generation and removal of isotopes in an MSR from fission, decay, activation and extraction. The extraction process was assumed to be mass-transfer limited, owing to the noble behavior and relatively low concentration of molybdenum in the reactor. It was found that the concentration at which Mo was present in the reactor had a profound effect on the specific activity of the Mo-99 that could be successfully extracted. If molybdenum is allowed to buildup in the reactor, the majority of Mo-99 is lost by decay or activation, leaving only a small fraction of Mo-99 in the extracted molybdenum. Therefore, separation processes that require a high relative molybdenum concentration for extraction are not desirable. On the other hand, separation processes become more difficult at low concentrations. Thus, a trade-off exists between the difficulty of the separation and effectiveness of the extraction as measured by the recovery fraction and product quality. It was found that increasing the power density enabled improved Mo-99 recovery for separation processes at increased concentrations. These results demonstrate the use of a model to explore the viability of Mo-99 extraction from an MSR, the influence of operating parameters on that extraction, and the conditions under which Mo-99 extraction can be optimized.

Abstract Under very low probability postulated accident like LOCA with unavailability of ECCS, the moderator heat sink maintains the channel integrity by providing adequate cooling. As the cooling is external to channel, the circumferential as well axial temperature gradient exists within the channel internals. An experimental set-up was fabricated to simulate the boil-off condition during a postulated accident in NPP. The experiments were conducted at average fuel bundle temperature 600 °C, 800 °C and 1100 °C with superheated mixture of steam and argon. The minimum and maximum temperature in CT, PT and fuel bundle were diametrically opposite to each other; the circumferential as well axial temperature gradient increased with an increase in bundle temperature. The rate of moderator heat absorption increases with increase in bundle temperature. The moderator was successfully able to prevent a breach in the channel; it proved to be the ultimate heat sink during a postulated LOCA with unavailability of ECCS.

Abstract In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The loads generated by the combustion process (increase of the containment atmosphere pressure and temperature) may threaten the integrity of the containment building and of internal walls and equipment. Turbulent deflagration flames may generate high pressure pulses, temperature peaks, shock waves and large pressure gradients which could severely damage specific containment components, internal walls and/or safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence. Currently, turbulence and steam effect on flame acceleration, flame deceleration and flame quenching mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose, two hydrogen deflagration benchmark exercises have been organised in the framework of the SARNET network. The first benchmark was focused on turbulence effect on flame propagation. For this purpose, three tests performed in the ENACCEF facility were considered. They concern vertical flame propagation in an initially homogenous mixture with 13 vol.% hydrogen content and different geometrical configurations. Three blockage ratios of 0, 0.33 and 0.6 were considered to generate different levels of turbulence. The second benchmark objective was the investigation of the diluting effect on flame propagation. Thus, three tests performed in the ENACCEF facility using the same blockage ratio of 0.63 and three different initial gas compositions (with 10, 20 and 30 vol.% diluents) have been considered. Since ENACCEF runs at ambient temperature, a surrogate to steam was used consisting of a mixture of 0.6He + 0.4CO 2 on molar basis. This paper aims to present the benchmarks conclusions regarding the ability of LP and CFD combustion models to predict the effect of turbulence and diluent on flame propagation.

Abstract In the present work the problem of the on-line monitoring of the reactivity in a source-free nuclear reactor is considered. The method is based on the classic point kinetic model of reactor physics. A relationship between the instantaneous value of the system stable period and the values of the neutron flux amplitude (or the power), of its derivative and of the integral convolution term determining the instantaneous value of the effective delayed neutron concentration is derived. The reactivity can then be evaluated through the application of the inhour equation, assuming the effective delayed neutron fraction and prompt generation time are known from independent measurements. Since the power related quantities can be assumed to be experimental observables at each instant, the reactivity can be easily reconstructed. The method is tested at first through the interpretation of power histories simulated by the solution of the point kinetic equations; the effect of the time interval between power detections on the accuracy is studied, proving the excellent performance of the procedure. The work includes also a study on the sensitivity of the reactivity forecast to the uncertainty on the values of the effective delayed neutron fraction and prompt generation time. The spatial effects are investigated by applying the method to the interpretation of flux evolution histories generated by a numerical code solving the space–time dependent neutron kinetic equations in the diffusion model. Also in this case the method proves to be quite effective in providing good estimates of the system reactivity, except at very short times after the introduction of a perturbation inducing a spatial transient. At last, the effect of the experimental noise is investigated, proving that the consequences in the accuracy of the reactivity prediction can be mitigated by using an adequate differentiation algorithm.

Abstract Differential neutron-scattering cross-sections of 165 Ho are measured at 40 or more scattering angles and at ≈0.5 MeV incident-energy intervals from ≈4.5–10.0 MeV. These results are combined with previously-reported neutron total and scattering cross-sections to construct a comprehensive experimental data base which is interpreted in the context of a coupled-channels model, with attention to the collective excitation of the K =7/2 − ground-state rotational band. Physical aspects are discussed and comparisons made with previous models reported in the literature and with the relevant portions of the ENDF/B-6 evaluated nuclear data file.