In recent years, global warming and climate change caused by the greenhouse gas emissions has given rise to widespread concerns. CO 2 has been considered as the principal greenhouse gas of interest, and fossil-fuel-fired power plants have been deemed as the largest stationary sources of CO 2 emission. It is imperative to capture CO 2 from these sources to reduce the global CO 2 emissions. Lately, capturing CO 2 from flue gas using solid absorbents shows promising for CO 2 abatement. For the cost-effective CO 2 capture process and the recycling of environmental pollutants, deprecated resources have been utilized for CO 2 capture from flue gas. In this work, fly ashes derived from different raw materials were tried as solid CO 2 sorbents for flue gas treatment. To improve their CO 2 capture capacities, the ashes were modified by different polyamines. An experimental demonstration on CO 2 capture behaviors of fresh ashes and modified sorbents in simulated flue gas atmosphere of 40°C, 15% CO 2 + 15% H 2 O and balanced N 2 was presented in detail with a fixed-bed reactor system. CO 2 capture capacities of fresh ashes were calculated as 0.56 mmolCO 2 /g, 0.32 mmolCO 2 /g, 0.44 mmolCO 2 /g and 0.83 mmolCO 2 /g, respectively. By contrast, CO 2 capture capacities of amine-modified samples had been enhanced as 0.38 mmolCO 2 /g, 0.65 mmolCO 2 /g, 1.07 mmolCO 2 /g, 0.85 mmolCO 2 /g and 1.17 mmolCO 2 /g. The optimal sample of TEPA-modified biomass ash (TEPA-BA) with CO 2 capture capacity of 1.17mmolCO 2 /g was screened. The optimal candidate was then selected for further investigation of the effects of temperature, CO 2 concentration and H 2 O concentration on its CO 2 capture behaviors. The results indicated that CO 2 capture capacity would increase with the increase of temperature in the range of 30 to 40 °C and decrease with the increase of temperature in the range of 40 to 60°C, increase with the increase of CO 2 concentration in the range of 5% to 20%, increase with the increase of H 2 O concentration in the range of 0% to 15% and decrease with the increase of H 2 O concentration in the range of 15% to 20%. The results in this work could provide basic data as a guidance for further applying the sorbents in practical operations.

Critical components of modern turbomachinery are frequently subjected to a myriad of service conditions that include diverse mechanical loads at elevated temperatures. The cost, applicability, and accuracy of either numerical or analytical component-level simulations are largely dependent on the material model chosen for the application. A non-interaction (NI) model derived from individual elastic, plastic, and creep components is developed in this study. The candidate material under examination for this application is 2.25Cr-1Mo, a low-alloy ferritic steel commonly used in chemical processing, nuclear reactors, pressure vessels, and power generation. Data acquired from literature over a range of temperatures up to 650°C are used to calibrate the creep and plastic components described using constitutive models generally native to general-purpose FEA. Traditional methods invoked to generate coefficients for advanced constitutive models such as non-linear kinematic hardening employ numerical fittings of hysteresis data, which result in values that are neither repeatable nor display reasonable temperature-dependence. By extrapolating simplifications commonly used for reduced-order model approximations, an extension utilizing only the cyclic Ramberg-Osgood coefficients has been developed to identify these parameters. Unit cell simulations are conducted to verify the accuracy of the approach. Results are compared with isothermal and non-isothermal literature data.

The concern about the global climate change and the unstable supply of fossil fuels stimulate the research of the new energy source utilization and the efficient energy system design. As the interests on the future energy sources and renovating the conventional power plants grow, an efficient and widely applicable power conversion system is required to satisfy both requirements. S-CO 2 cycle is considered as a promising candidate with the advantages of 1) relatively high efficiency in the modest temperature (450–750°C) region because of non-ideal properties near the critical point, 2) effectively reduced size of the total cycle with compact turbo-machines and heat exchangers, 3) potential for using in various applications with competitive efficiency and simple layout. The S-CO 2 cycle was originally considered as an attractive candidate for the power conversion cycle of the next generation nuclear reactors. However, due to many benefits of the S-CO 2 cycle, it is recently considered in other conventional and renewable energy system applications including fossil fuel power plant system, ship propulsion application, concentrated solar power system, fuel cell bottoming power cycle and so on. This paper will discuss about the design of S-CO 2 cycle for the various energy system applications over different temperature range. Unlike a large size power plant which usually focuses more on maximizing the cycle efficiency, a small capacity energy system is seriously concerned about the total size of the cycle. In this manner, several preliminary S-CO 2 cycle designs will be compared in terms of the efficiency and the physical size. Various layouts and components of S-CO 2 cycle are compared to find the optimum cycle for each energy systems. The in-house codes developed by the KAIST research team are used to evaluate the various cycle performances and component preliminary designs. The obtained results will be compared to the conventional power conversion systems along with its implication to other existing designs.

Since the beginning of the twenty-first century, energy conservation has become a major feature of interest in most industrialised countries. The economics of saving energy versus wasting it has driven industrial activists to pay more attention to energy conservation. The implementation of energy conservation requires that all the possibilities of counteracting any potential loss of energy must be considered. This includes reducing heat losses from furnaces, thermal insulation, repair of steam leaks in power plants, heat loss from nuclear reactors, and all other practices that may be implemented rapidly and, preferably at low cost. Once this is achieved, further strategies have to be developed to stabilise short-term energy conservation in systems by implementing permanent solutions. Permanent energy conservation solutions are more expensive, but result in energy benefits over many years. These permanent solutions are referred to as Waste heat recovery systems (WHRSs). This paper presents potential application of WHRSs in high-temperature reactors technology. WHRSs have attracted the attention of many researchers over the past two decades, as using waste heat improves the system overall efficiency, notwithstanding the cost of extra plant. WHRSs require specially designed heat recovery equipment, and as such the used and/or spent HTR fuel tanks were considered by the way of example. An appropriately scaled system was designed, constructed and tested to demonstrate the functioning of such a cooling system first and validated the theoretical model that simulates the heat transfer process in the as-designed WHRS. It is a one-dimensional flow model assuming quasi-static and incompressible liquid and vapour flow its mathematical simulations as developed in Part I (Senda and Dobson, 2013). Two separate and independent cooling lines, using natural circulation flow in a particular form of heat pipes called thermosyphon loops were used to ensure that the fuel tank (FT) is cooled when the power conversion unit has to be switched off for maintenance, or if it fails.

GTHTR300C is a small modular reactor based on a 600 MWt high temperature gas reactor (HTGR) and intended for a number of cogeneration applications such as process heat supply, hydrogen production, steelmaking, desalination in addition to power generation. The basic design has been completed by JAEA together with Japanese heavy industries. The reactor design and key plant technologies have been validated through test reactor and equipment verification. Future development includes demonstration programs to be performed on a 50 MWt system HTR50S. The demonstration programs are implemented in three steps. In the first step, a base commercial plant for heat and power is to be constructed of the same fuel proven in JAEA’s successful 950°C, 30 MWt HTGR test reactor and a conventional steam turbine such that the construction can readily proceed without major development requirement and risk. Beginning in the second step, a new fuel presently being developed at JAEA is expected to be available. With this fuel, the core outlet temperature is raised to 900°C for purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. Added in the final step is a thermochemical process to demonstrate nuclear-heated hydrogen production via water decomposition. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The designs of GTHTR300C and HTR50S will be presented and the demonstration programs will be described.

In a nuclear power plant, one of the important issues is evaluation of the safety of reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs. However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there are little knowledge how coolant in a core response when large earthquake acceleration is added. Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction or the power of core is fluctuated with the oscillation by the experiments and numerical analysis. However detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow structure. In order to investigate it, we visualize changing of bubbly flow structure in pipeline on which sine wave is applied. Bubbly flow is produced with injecting gas into liquid flow through a horizontally circular pipe. In order to vibrate the test section, the oscillating table is used. The frequency of vibration added from the table is from 1.0 Hz to 10 Hz and acceleration is from 0.4 G to 1 G (1 G = 9.8 m/s 2 ). The test section and a high speed video camera are fixed on the table. Thus the relative velocity between the camera and the test section is ignored. In the visualization experiment, the PIV measurement is conducted. Then the motion of bubbles, for example the shape, the positions and the velocity are measured with observation. In addition, by varying added oscillation amplitude, frequency and flow rate of the fluids, the correlation between these parameters and bubble motion was evaluated. It was clarified that the behavior of liquid phase and bubble through horizontal circular pipes was affected by an oscillation. When structure vibration affects the flow, two main mechanisms are supposed. One is the addition of body force of the oscillation acceleration to liquid phase and bubble, and the other is the velocity oscillation of the test section and the effect of the boundary layer of the pipe wall. It was also found that when the added oscillation frequency and amplitude was changed, the degree of the fluctuation of liquid phase and bubble motions were changed.

In order to ensure the compactness and high-power density of a nuclear power reactor, the research on tight-lattice fuel bundle, with a narrow gap distance between fuels, has been highlighted. Recently, KAERI (Korea Atomic Energy Research Institute) has been developing dual-cooled annular fuel to increase a significant amount of the reactor power in OPR1000 (Optimized Power Reactor), a PWR (Pressurized Water Reactor) optimized in the Republic of Korea. The dual-cooled annular fuel is configured to allow a coolant flow through the inner channel as well as the outer channel. To introduce the dual-cooled annular fuel to OPR1000 is aiming at increasing the reactor power by 20% and reducing the fuel-pellet temperature by 30%, as compared to the cylindrical solid fuel, without a change in reactor components. In such a case, due to larger outer diameter of a dual-cooled annular fuel, the dual-cooled annular fuel assembly exhibits a smaller P/D (Pitch-to-Diameter ratio) than the conventional cylindrical solid fuel assembly. In other words, the dual-cooled annular fuel array becomes the tight-lattice fuel bundle configuration, and such a change in P/D can significantly affect the thermal-hydraulic characteristics in nuclear reactor core. In this paper, the pressure drop and flow pulsation in tight-lattice rod bundle were investigated. As the test sections, the tight-lattice rod bundle of P/D = 1.08 was prepared with the regular rod bundle of P/D = 1.35. The friction factors in P/D = 1.08 appeared smaller than those in P/D = 1.35. For P/D = 1.08, the twist-vane spacer grid became the larger pressure loss coefficients than the plain spacer grid. In P/D = 1.08, the flow pulsation, quasi-periodic oscillating flow motion, was visualized successfully by PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques. The peak frequency and power spectral density of flow pulsation increased with increasing the Reynolds number. Our belief is that this work can contribute to a design of nuclear reactor with tight-lattice fuel bundle for compactness and power-uprate and a further understanding of the coolant mixing phenomena in a nuclear fuel assembly.

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors were contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void faction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also body force on the two-phase flow and shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake. Therefore, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in a series of study. In this study, to develop the prediction technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT was expanded to two-phase flow simulation under earthquake conditions. In this paper, outline of expansion of the TPFIT to simulate detailed two-phase flow behavior under the earthquake condition was shown. And the bubbly flow in a horizontal pipe excited by oscillation acceleration and under the fluctuation of the liquid flow was simulated by using expanded TPFIT. Predicted deformation of bubbles near wall was compared with measured results under flow rate fluctuation and structural vibration.

In order to investigate the possible effect of seismic vibration on two-phase flow dynamics and thermal-hydraulics of a nuclear reactor, experimental tests of adiabatic air-water two-phase flow under low-frequency vibration were carried out in this study. An eccentric cam vibration module operated at low motor speed (up to 390rpm) was attached to an annulus test section which was scaled down from a prototypic BWR fuel assembly sub-channel. The inner and outer diameters of the annulus are 19.1mm and 38.1mm, respectively. The two-phase flow operating conditions cover the ranges of 0.03≤< j g > ≤1.46m/s and 0.25≤< j f >≤1.00m/s and the vibration displacement ranges from ±0.8mm to ±22.2mm. Steady-state area-averaged instantaneous and time-averaged void fraction was recorded and analyzed in stationary and vibration experiments. A neural network flow regime identification technique and fast Fourier transformation (FFT) analysis were introduced to analyze the flow regimes and void signals under stationary and vibration conditions. Experimental results reveal possible changes in flow regimes under specific flow and vibration conditions. In addition, the instantaneous void fraction signals were affected and shown by FFT analysis. Possible reasons for the changes include the applied high acceleration and/or induced resonance at certain ports under the specific flow and vibration conditions.

Under certain conditions, boiling water reactors (BWRs) would be susceptible to couple neutron-thermalhydraulic instability. It is important to predict such potential problems as early as possible and prevent the core instability from happening. In each BWR reload core design, fuel vendors are required to provide instability boundaries on power/flow map to assure safety operation of the nuclear reactor. In Taiwan, a LAPUR5.2 methodology had been adapted to build up the remarkable analysis mode for different types BWRs to verify vendor’s results. However, with upgrading nuclear safety technology, most of boiling water reactors has been adopting partial length fuel assemblies to reduce two-phase pressure drop and void fraction, to improve reactor stability. The question is that LAPUR5.2 methodology cannot precisely analysis stability characteristics from the variation of flow area in fuel assemblies. From the reasons of upgrading stability analysis, a LAPUR6.0 methodology had built to do the related researches. This research was based on a comparison study between LAPUR5.2 and LAPUR6.0 to realize the major differences and their effects on stability characteristics. According to the comparison results for Kuosheng Nuclear Power Plant Unit 2 Cycle 21 reload design, it shows that LAPUR6.0 can completely present pressure drop, void fraction and density reactivity coefficient from the changing of flow area and fuel spacers.

The Darlington New Nuclear Project (DNNP) is a proposal by Ontario Power Generation for the site preparation, construction, and operation of up to four new nuclear reactor units for the production of up to 4800 MW of electrical generating capacity. The site was selected by the Ontario Government and is located at the existing Darlington Nuclear site, which is located on the north shore of Lake Ontario and about 70 km east of Toronto, Canada. The regulatory process for the project began in September 2006 and has included the completion of a comprehensive Environmental Assessment (EA), the submission of a Licence to Prepare Site Application, and a three-week public hearing from March 21 to April 8, 2011. This paper provides an overview of various site evaluation and safety studies that were performed in order to demonstrate that the DNNP site meets the Canadian regulatory requirements. The site evaluation studies are also consistent with the principles in the IAEA document NS-R-3, “ Site Evaluation for Nuclear Installations ” and its associated guides. Accordingly, the site evaluation studies considered the following hazards: extreme meteorological events, flooding hazards, seismic hazards, geotechnical hazards, external human-induced events, and potential dispersion of radioactive material with off-site dose consequences. These hazards were assessed in terms of risk to the new nuclear units and ultimately to the public and the environment. Since a reactor technology has not yet been selected for the DNNP, a multi-technology approach was followed for both the EA and site evaluation studies. This involved the use of a bounding Plant Parameters Envelope (PPE), similar to the US-based PPE approach, encompassing the following reactor designs: the US EPR (1580 MWe), the AP1000 (1037 MWe), the ACR-1000 (1085 MWe), and the Enhanced CANDU 6 (740 MWe). Additionally, to assess the impact of protective measures on the local population (e.g., in terms of temporary evacuation), bounding source terms were derived based on the regulatory safety goals for both Small Release Frequency and Large Release Frequency. These generic source terms are expected to bound the releases from any credible accidents, for any reactor designs considered licensable in Canada. In each of the hazard areas, the risk was determined to be acceptably low or could be reduced to an acceptable level through design mitigation. The overall conclusion is that the DNNP site is suitable for the new nuclear units.

Leaching yields from an activated aggregate, removal yields from leachate, and overall mass balance relevant to the chemical procedure were investigated on radionuclides of 59 Fe, 60 Co, and 152 Eu to reduce volume of radioactive waste used in a nuclear reactor facility. The radioactive nuclides were leached with 7 M HNO 3 from the aggregate, then the 7 M HNO 3 solution was mixed with NaOH or NH 3 water to remove the nuclides as a precipitate. 70–80% of 60 Co and 60–70% of 152 Eu were leached from an activated aggregate. 100% of 59 Fe, more than 99% of 60 Co, and 94–98 % of 152 Eu were removed as a precipitate at pH higher than 4, 7, and 6, respectively. Removal rates became lower at pH lower than 4 for 59 Fe, 7 for 60 Co, and 6 for 152 Eu. It is basically possible to separate 60 Co and 152 Eu from precipitate of Fe at pH ∼ 4. The further separation to reduce volume of waste is not practical, because 20% of 60 Co and 152 Eu contained in the precipitate. 93% of volume reduction was achieved with one-step separation.

10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, is a voluntary regulation that allows nuclear power plants to categorize structures, systems and components (SSCs) according to their risk-informed safety significance and then adjust the treatment applied to these SSCs commensurate with their safety significance. 10 CFR 50.69 allows the reduction of many special treatment requirements for safety related SSCs that have been categorized as low safety significant in accordance with an approved methodology. Conversely, SSCs that are non safety related but categorized as safety significant are required to be evaluated for additional controls to ensure that they can perform their safety significant functions. 10CFR 50.69 does not alter the design requirements and safety classification of any categorized component. The intent of 50.69 is to allow increased focus and resources to be applied to safety significant SSCs while allowing increased flexibility for items with low safety significance. The expected result of applying this methodology is that nuclear safety is increased while allowing the reliable and cost-effective operation of the power plant. To date, none of the utilities in the USA have fully implemented 10 CFR 50.69 since it became effective. Two plants, Wolf Creek and Surry, completed categorization of select systems. However, they never submitted a License Amendment Request (LAR). With increased focus on risk-informed regulation, NRC is interested in seeing to see a utility implement 50.69. The Vogtle Electric Generating Plant (VEGP) initiated the “50.69 Project” at the beginning of 2011. The ultimate goal is to seek approval from NRC (via license amendment) to fully implement 10 CFR 50.69. The project has three high level milestones — prepare and submit LAR; NRC review and approval; and implement 10 CFR 50.69. VEGP applied for the pilot status on December 6, 2010, and received it on June 17, 2011. This paper outlines lessons learned and challenges faced year-to-date while performing engineering work to achieve the first milestone. The lessons learned can be used by any licensees interested in pursuing 10 CFR 50.69 at their sites before starting the 50.69 project.

One of the important parameters that affect the thermal-hydraulic performance of nuclear fuel assemblies is the spatial — that is the lateral and the axial — distribution of power. Since this parameter may have a significant influence on thermal margins of nuclear reactors, it is necessary to take it into account in various models and/or correlations. One practical difficulty in doing so is the fact that the spatial power distribution is a function of space variables, which makes it very inconvenient to implement into single-parameter correlations. In addition, there is still lack of a simple theoretical model that captures the effect of spatial power distributions on the thermal-hydraulic performance of nuclear fuel assemblies. In this paper, an accurate and fast running convolution method is presented to predict the influence of axial power distribution on wall temperature distributions. The method has been verified against CFD predictions of the wall temperature in a heated pipe and an excellent agreement between the two approaches is demonstrated. The method is applicable only for constant fluid properties and for a fully developed flow regime.

In the context of a plant modernization, developing digital I&C technology is a crucial challenge to improve nuclear plants safety and reliability. Digital technology is usually oriented to achieve functions such as plant control, monitoring, simulation and protection in a user-friendly way. On the other hand, the analogue instrumentation implemented in the so-called “old generation consoles” is often essential and not immediately or completely replaceable. As a consequence, the interaction between the analogue and digital data seems to be a necessary step before starting the digital I&C licensing process. The fundamental difference between analogue and digital technologies relies on the fact that digital logic is based on processors, hence it can be customized by programming its software. However, introducing new code can result in a new set of potential failure modes to be accounted for. As a consequence, original analogue systems mostly assure a higher level of protection with respect to digital systems. In this scenario, a benefit could arise from the use of Field-Programmable Gate Arrays (FPGAs), based on a hardware architecture whose routing is made via software, thus resulting in a variety of possible tasks. FPGAs’ employment ranges from automotive and industrial applications, ASIC prototyping, software defined radios, radar, image and DSP. In this work a critical analysis of FPGA fundamental features and potentialities in nuclear plant I&C design is achieved in conjunction with some practical applications. Troubles arising from coping with processor-based system are presented and compared to benefits and potentialities offered by FPGA real-time architectures: indeed, FPGAs comprise a higher number of logic blocks and functions able to manage parallel processes with self triggering, and provided into a “non-frozen” structure but easily reconfigurable. This characteristics of being in-system programmable (ISP), i.e. a device capable of being programmed while remaining resident in a high-level system, can be considered as the main advantage of using FPGA. The employment on a large scale is also justified by its high determinism and testability, leading to high performance in terms of reliability. As a case-study, we propose a supervisory full-digital system that has been designed, realized, tested and validated implementing a FPGA architecture to be used in parallel to the TRIGA nuclear reactor RC-1 analogue console at the ENEA Casaccia Research Centre in Rome. We report on the design choices, and on pros and cons of using FPGA instead of the classical processor-based architectures. This preliminary apparatus has been developed using the LABVIEW environment and FPGA-based technology, an appropriate tool to get across simulation to Hardware-in-the-Loop (HIL) technique allowing to move on production from prototyping.

Energy utilization from low-grade fuels of either fossil or renewable origin, or from medium-temperature heat sources such as solar, industrial waste heat, or small nuclear reactors, for small-scale power generation via steam cycles, can be reasonably enhanced by a simple technology shift. This study evaluates the technical feasibility of a compact power generation package comprising a steam turbine directly coupled to a high-speed alternator delivering around 8–12 MW of electrical power. Commercial or research-phase high-speed electrical generators at MW-scale are reviewed, and a basic thermodynamic design and flow-path analysis of a steam turbine able to drive such a generator is attempted. High-speed direct drives are winning new grounds due to their abilities to be speed-controlled and to avoid the gearbox otherwise typical for small system drivetrains. These two features may offer a reasonable advantage to conventional drives in terms of higher reliability and better economy. High-speed alternators with related power electronics are nowadays becoming increasingly available for the MW-size market. A generic 8 to 12 MW synchronous alternator running respectively at 15,000 to 10,000 rpm, have been used as a reference for evaluating the fundamental design of a directly coupled steam turbine prime mover. The moderate steam parameter concept suits well for converting mid-temperature thermal energy into electrical power with the help of low-tech steam cycles, allowing for distributed electricity production at reasonable costs and efficiency. Steam superheat temperatures below 350°C (660°F) at pressures of maximum 20 bar would keep the steam volumetric flow sufficiently high in order to restrain the turbine losses typical for small-scale turbines, while helping also with simpler certification and safety procedures and using primarily established technology and standard components. The proposed steam turbines designs and their characteristics thereof have been evaluated by computer simulations using the in-house code ProSteam and its sub-procedures AXIAL and VaxCalc, by courtesy of Siemens Industrial Turbomachinery and its steam turbine division located in Finspong, Sweden. The first results from this study show that high-speed steam turbines of the proposed size and type are possible to design and manufacture based on conventional components, and can be expected to deliver a very satisfactory performance at variable power output.

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

Hydrogen accumulation at the top of the containment or reactor building may occur due to an interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Dai-ichi). Hydrogen accumulates usually in a containment of nuclear reactor as a stratified semi-confined layer of hydrogen-air mixture. Detonation of such mixture may lead to significant damage of the containment structure. A series of large scale experiments on hydrogen combustion and detonation in a semi-confined layer of uniform and non-uniform hydrogen-air mixtures in presence of obstructions or without them was performed at the Karlsruhe Institute of Technology (KIT). Critical conditions for deflagration-to-detonation transition and then for steady state detonation propagation were experimentally evaluated in a flat semi-confined layer. The experiments were performed in a horizontal semi-confined layer with dimensions of 9×3×0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 0–34 vol.% without or with a gradient of 0–1.1 mol. %H 2 /cm. Effects of hydrogen concentration gradient, thickness of the layer, geometry of the obstructions, average and maximum hydrogen concentration on critical conditions for detonation onset and then propagation were investigated with respect to the safety analysis. Blast wave strength and mechanical response of the safety volume was experimentally measured as well.

Since 1999 ENEA is developing the heavy liquid metal (HLM) technology aiming to support the design and implementation of a Lead cooled Fast Reactor (LFR) and an Accelerator Driven System (ADS), both in the frame of the Italian and European research programs. In these contexts several experiments have been performed, in different fields, going from coolant thermal-hydraulic, component development and structural material characterization. Recently, in the frame of the IP-EUROTRANS (6 th Framework Program EU), domain DEMETRA, ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of a pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, named “Integral Circulation Experiment (ICE)”, has been implemented thanks a joint effort of several research institutes, mainly ENEA and University of Pisa, allowing to design an appropriate test section. This has been installed in the CIRCE facility, the largest worldwide experimental facility for the HLM technology investigation. The goal of the experiments was to demonstrate the technological feasibility of a heavy liquid metal (HLM) pooltype nuclear system in a relevant scale (1 MW), investigating the related thermal–hydraulic behavior under both steady state and transient conditions. This paper reports a description of the experiment, as well as the results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test, an un-protected loss of heat sink (ULOH) test, and an un-protected loss of flow (ULOF) test. The post-test analyses of the experiments is presented. The whole domain has been modeled by a suitable 1-D nodalization, and the results carried out have been studied performing numerical calculations by the REALP5 system code modified to take in account the LBE thermal-physical properties when employed as nuclear coolant. The obtained experimental results as well as the performed post-test analysis have demonstrated the thermal-hydraulic and technological feasibility of a pool-type nuclear system cooled by HLM.

Safety evaluation is one of the most important aspects of the nuclear reactors. Here some safety studies have been carried out on High Temperature Gas cooled Reactors (HTGR) with fuel in the form of pebbles. This is one of the six types of GEN IV reactors that are being developed internationally with emphasis on features like inherent safety, nuclear proliferation resistance and high thermal to electric conversion efficiency. This work pertains to the dynamic simulation of the reactor with, as well as without reactivity feedback. It involves the solution of point kinetics equations with reactivity feedback arising from the power rise. First, the reactivity feedback arising from the power rise is calculated in the form of power coefficient of reactivity. The steady state temperatures are calculated at two different steady state power levels of the reactor using the heat balance between the fission power produced in the reactor and the heat removed by the coolant. A simplified model of the reactor and lumped model of heat transfer is developed and used. The temperature rise in going from one power level to other power level is calculated. This is multiplied by typical value of the temperature coefficient of reactivity available in the literature. The reactivity changes divided by the power rise provides the power coefficient of reactivity. For performing dynamic analysis of the reactor, the kinetics equations along with feedback are solved. The methodology is verified against the analytical expressions available. Simulations are carried out for the case of raising the power of the reactor from low power to high power and transients at full power and shutting down of the reactor by reactor SCRAM. It is observed that for raising the power of the reactor, the reactivity addition beyond 20pcm/second may not be acceptable as the higher reactivity addition rates result in lower instantaneous reactor period resulting in SCRAM of the reactor. This information is useful in designing the short term and long term decay heat power removal system. Moreover, it provides insight on the determination of the maximum permissible reactivity addition rates in the reactor and optimization of the reactivity feedback due to rise in power.