Generation Of Nuclear Power Plants Research Articles

Application of Modified CANDLE Burnup to Very Small Long Life Gas-Cooled Fast Reactor

Gas-cooled Fast Reactor is a good candidate for fourth generation nuclear power plant that projected to be used started in 2030. In this study, modified CANDLE burn-up strategy is adopted to create 300 MWt long life Gas-cooled Fast Reactor with metallic fuel U-10wt%Zr without enrichment. This design demonstrated excellent performance with the average discharge burn-up is about 25.9% HM.

Effect of shear connectors on local buckling and composite action in steel concrete composite walls

Steel concrete composite (SC) walls are being used for the third generation nuclear power plants, and also being considered for small modular reactors. SC walls consist of thick concrete walls with exterior steel faceplates serving as reinforcement. These steel faceplates are anchored to the concrete infill using shear connectors, for example, headed steel studs. The steel faceplate thickness (tp) and yield stress (Fy), and the shear connector spacing (s), stiffness (ks), and strength (Qn) determine: (a) the level of composite action between the steel plates and the concrete infill, (b) the development length of steel faceplates, and (c) the local buckling of the steel faceplates. Thus, the shear connectors have a significant influence on the behavior of composite SC walls, and should be designed accordingly. This paper presents the effects of shear connector design on the level of composite action and development length of steel faceplates in SC walls. The maximum steel plate slenderness, i.e., ratio of shear connector spacing-to-plate thickness (s/tp) ratio to prevent local buckling before yielding is also developed based on the existing experimental database and additional numerical analysis.

Preparation of 240Pu and 242Pu targets to improve cross-section measurements for advanced reactors and fuel cycles

240Pu and 242Pu targets were prepared by the so-called “molecular plating” on an Al backing for cross-section measurements. The total activity of the actinide layer was determined by low-geometry alpha-particle counting. The atomic abundances of the two Pu materials were determined via thermal ionization mass spectrometry. A radiochemical separation was performed by anion exchange prior to the preparation of the layers to prevent interferences with the ingrown daughter nuclide during the characterization and cross-section measurements. The targets were prepared to be used in the project “Metrology for New Generation Nuclear Power Plants” (MetroFission), within the frame of the European Metrology Research Programme, directed to improve the knowledge on neutron cross sections through metrological approaches. For measurements of the 240Pu(n,f) and 242Pu(n,f) cross sections 240Pu and 242Pu targets were produced and mounted in fission chambers. In parallel, 240Pu and 242Pu targets were prepared for the ANDES project (Accurate Nuclear Data for Nuclear Energy Sustainability) for cross-section measurements at the CERN neutron time-of-flight facility n_TOF, and the Van de Graaff laboratories of IRMM and CNRS/CENBG.

SEM analysis of ion implanted SiC

SiC is a material used in two future energy production technologies, firstly as a photovoltaic layer to harness the UV spectrum in high efficient power solar cells, and secondly as a diffusion barrier material for radioactive fission products in the fuel elements of the next generation of nuclear power plants. For both applications, there is an interest in the implantation of reactive and non-reactive ions into SiC and their effects on the properties of the SiC. In this study 360keV Ag+, I+ and Xe+ ions were separately implanted into 6H–SiC and in polycrystalline SiC at various substrate temperatures. The implanted samples were also annealed in vacuum at temperatures ranging from 900°C to 1600°C for various times. In recent years, there had been significant advances in scanning electron microscopy (SEM) with the introduction of an in-lens detector combined with field emission electron guns. This allows defects in solids, such as radiation damage created by the implanted ions, to be detected with SEM. Cross-sectional SEM images of 6H–SiC wafers implanted with 360keV Ag+ ions at room temperature and at 600°C and then vacuum annealed at different temperatures revealed the implanted layers and their thicknesses. A similar result is shown of 360keV I+ ions implanted at 600°C into 6H–SiC and annealed at 1600°C. The 6H–SiC is not amorphized but remained crystalline when implanting at 600°C. There are differences in the microstructure of 6H–SiC implanted with silver at the two temperatures as well as with reactive iodine ions. Voids (bubbles) are created in the implanted layers into which the precipitation of silver and iodine can occur after annealing of the samples. The crystallinity of the substrate via implantation temperature caused differences in the distribution and size of the voids. Implantation of xenon ions in polycrystalline SiC at 350°C does not amorphize the substrate as is the case with room temperature heavy ion bombardment. Subsequent annealing of the implanted polycrystalline samples leads to increased thermal etching effects such as grain boundary grooving. Damage due to channelling (or non-channelling) in the different crystallites resulted also in differences in thermal etching in the crystallites.

Research on Constant Velocity Extruding Process Control for 36,000-Ton Vertical Extrusion Press

Metal extrusion is one of the most significant methods in the field of plastic deformation. The 36,000-ton vertical extrusion press uses one-piece hot extrusion molding method to produce high-pressure, large-diameter, thick-walled, seamless, alloy-steel pipe, which is required by ultrasupercritical power generator units and the third generation of nuclear power plants. The main table (extruding platform) is driven by 36,000-ton thrust. The extrusion process is a typical nonlinear multivariable strong-coupling finite-time system. In this paper, we analyze and solve the large-scaled engineering control problems, including (1) discussion of the mechanical structure of the controlled object and the feature of multivariable strongly coupling for large-scaled parallel-driving hydraulic system, (2) research on the engineering practical simplified control algorithm for multivariable system and propose a way to solve the coupling problem of the hydraulic system, and (3) design of double closed-loop control of velocity and pressure for the new technology of constant velocity extrusion. In our paper, we have proven through practice that the traditional proportional intergral (PI) controlling method, with proper controlling strategy, is still the most efficient and practicable way for a large-dimension complex object in industry.

Mechanism of plastic deformation of a Ni-based superalloy for VHTR applications

Alloy 617 is considered a leading structural material for next generation nuclear power plants due to its good corrosion resistance and exceptional strength at high-temperatures. In the present work, a complementary study of deformation mechanisms in Alloy 617 was performed via tensile testing at temperatures up to 1000°C. Dynamic strain aging characterized by serrated flow leading to temperature independent yield strength was observed at intermediate temperatures. At temperatures higher than 700°C, the material’s strength reduced dramatically as a result of the predominance of dislocation creep and dynamic recrystallization. The changes in deformation mechanisms were established by direct observations of Electron Backscatter Diffraction mappings and by the analysis of texture development based on the pole figure mappings. Recrystallized grains were found to preferentially grow in particle-rich areas and on high-angle grain boundaries. Finally, fracture was initiated from particle cracks at temperatures up to 700°C and by triple junction cracks from 800 to 1000°C.

다중화 구조 제어시스템에 대한 신뢰도 분석

Since a slight malfunction of control systems in a nuclear power plant may cause huge catastrophes, such control systems usually have multiple redundancy and reliable features, and their reliability and availability should be analyzed and verified thoroughly. This paper performed the reliability analysis of the SPLC (Safety Programmable Logic Controller) that is under developed as the control systems for the next generation nuclear power plant. One of the key features of SPLC is that it has multiple redundancy modes as faults happen, which means the reliability analysis for one fixed redundant model is not enough to analyze the reliability of SPLC. With considering this reconfigurable concept, FTA (Fault Tree Analysis) was used to capture fault-relationship among sub-modules. The analysis results show that MTTF (Mean Time to Fault) of SPLC is 45,080 hours, which is a about 4.5 times longer than the regulation, 10,000 hours.

Actual status of the research alliances “Condensation Induced Water Hammer” and “Boiling processes in Pressurized Water Reactors”

The German Nuclear Society (KTG) is a non-profit scientific technical association with more than 2500 members. KTG strives for the development and dissemination of the stateof-the-art in nuclear science and technology to promote the peaceful use of nuclear energy and related disciplines. The technical work of the KTG members is performed within the framework of currently 9 topical sections, which, for instance, jointly organize the KTG Annual Meetings, topical meetings devoted to selected subjects and several other activities. In recent years, the topical sections Reactor Safety, Thermal and Fluid Dynamics, as well as Reactor Physics and Computational Methods, have successfully established a series of meetings on \Topical Issues of Reactor Safety Research in Germany. These meetings represent a forum for authorities, operators, technical support organizations, universities and vendors. A Technical Program Committee (TPC) consisting from members of the participating topical sections is responsible for selecting the issues to be addressed at the meeting. The acceptance and success of this series of meetings is mainly due to the current subjects dealt with. In addition to the presentation of scientific and technical research results, this forum offers the chance for a free exchange of opinions outside licensing and supervisory procedures. A balance is presented between survey papers on the background and the state of the art in the related field, specific technical presentations, and outlook papers. The participants are provided with a general overview on the area addressed and with new ideas, and impetus for their future work. In order to further promote these conference series, the TPC has initiated co-operation with the journal KERNTECHNIK. The cooperation started in 2011 with a very successful edition of the topic \Requirements for the demonstration of effective emergency core cooling during loss-ofcoolant accidents involving the release of insulation material and other substances. Two contributions of this special issue were among the top five downloads in the period 2010 up to 2011 of KERNTECHNIK. The 2012 meeting on \Topical Issues of Reactor Safety Research in was held at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) on 11–12 October 2012, hosted by the Institute of Fluid Dynamics of HZDR and the TUV NORD SysTec GmbH & Co. KG and the section Thermal and Fluid Dynamics of the German Nuclear Society (KTG). At the meeting selected results of the two research alliances \Condensation induced water hammer – CIWA and \Boiling processes in PWR were presented. Both research alliances are sponsored by the Federal Ministry of Education and Research within the Energy Research 2020+ concept. Even though Germany is phasing out of domestic power generation from nuclear energy, it should be noted that the Federal Ministry of Education and Research continues to fund further scientific and technological R&D in support of safe operation of existing nuclear reactors. One of the most effective paths for ensuring high quality basic and applied research programs is to provide funding of multi-organizational, multi-disciplinary research alliances. In addition, special attention should be paid to the support of young scientists and researchers, and to the dissemination and transfer of research findings to industry. The selection of the main topics \Condensation Induced Water Hammer and \Boiling processes in PWR at the actual topical meeting was made by the following background: Pressure surges are still a cause of serious damages in both nuclear and conventional power generation plants. Understanding the mechanisms and underlying physics, as well as developing measures to prevent and/or mitigate their initiations are still challenging tasks for utilities and technical support organizations. Boiling processes play an important role in energy generating facilities were energy transfer is essential. The deeper understanding of the mechanisms not only increases the safety of the facility but also enables to enhance the effective operation. In the past the development of promising Computational Fluid Dynamics (CFD) models stagnated not least due to missing of CFD grade experiments. The research alliance \Boiling processes in PWR contributed in both directions: Small and large scale experiments and model development. For enhanced dissemination of the state-of-theart, selected papers of the meeting are published in this special issue of KERNTECHNIK. Both, the topical meeting as well as the now presented special issue of KERNTECHNIK is certainly an impressive proof that the strategic objectives of the Basic Energy Research 2020+ concept can be achieved. EDITORIAL

「震災・原発事故と日本農業の国際化対応」

「震災・原発事故と日本農業の国際化対応」

A T-H Model for Residual Heat Removal System Reliability Analysis

Passive system is widely used in new generation nuclear power plant (NPP) such as AP1000 and high-temperature gas-cooled reactor in order to improve the safety of NPP. Since system operation relies on natural force, and both the driven force and resistance are influenced by many uncertain factors, physical process failure becomes one of the significant contributors to the system operation failure, which need to be considered in system reliability. For system physical process failure analysis, the thermal-hydraulic (T-H) model should be established to describe the system performance, namely the system can operate successfully or not. And the T-H model should be fast enough to be used in the system reliability analysis,In this paper, the T-H model of passive residual heat removal system (PRHR) in the AP1000 for analysing the system reliability is described. The accuracy and the validity of the model have been proved by comparing the results with the requirements of AP1000 design. Finally, the connection between residual power and flow rate in AP1000 PRHRS is analysed.

The Application of Cognitive Interface Design Methodology for a Digitalized Main Control Room in Nuclear Power Plant

This technical study describes the application of Ecological Interface Design (EID) method that may improve operator performance in digitalized system environment such as main control room in next generation nuclear power plant. While cognitive interface design method such as Information Rich Display (IRD) is mainly focused on stereotyped information for the user, EID method helps users' mental resources to be allocated to tasks that process highly abstract information including diagnosis and situation assessment. However, the sophisticated and relationship-embedded EID displays may cease to be intuitive when many of them are presented without consistently designed common elements. This study aims at suggesting consistent elements of EID displays for a digitalized Main Control Room of Nuclear Power Plant taking the example of the control process of Steam Generator (SG) Water Level. The proposed EID displays were reviewed by domain experts and operators to conjecture how such EID displays might benefit the operator's decision making. The experience from this study indicates that situation awareness in digitalized interface design may be improved by appropriately designed visualization of abstract information and that operator may be aided by the represented dynamics in EID display.

Defect microstructural evolution in ion irradiated metallic nanofoils: Kinetic Monte Carlo simulation versus cluster dynamics modeling and in situ transmission electron microscopy experiments

Understanding materials degradation under intense irradiation is important for the development of next generation nuclear power plants. Here we demonstrate that defect microstructural evolution in molybdenum nanofoils in situ irradiated and observed on a transmission electron microscope can be reproduced with high fidelity using an object kinetic Monte Carlo (OKMC) simulation technique. Main characteristics of defect evolution predicted by OKMC, namely, defect density and size distribution as functions of foil thickness, ion fluence and flux, are in excellent agreement with those obtained from the in situ experiments and from previous continuum-based cluster dynamics modeling. The combination of advanced in situ experiments and high performance computer simulation/modeling is a unique tool to validate physical assumptions/mechanisms regarding materials response to irradiation, and to achieve the predictive power for materials stability and safety in nuclear facilities.

Implementation of heat production and storage technology and devices in power systems

Implementation of heat storage devices and technologies at power generation plants is a promising way to provide more efficient use of natural energy resources. Heat storage devices can partly replace conventional heating technologies (such as direct use of fossil fuels) during peak energy demand or in the situations where heat and electricity supply and demand do not coincide and to obtain low cost heat energy which can be further transmitted to industrial, commercial and domestic consumers.This paper presents the innovative Heat Production and Storage Device and its application at conventional, nuclear and renewable power generation plants for optimization and balancing of electricity grids. The Heat Production and Storage Device is a vessel type induction-immersion heat production and storage device which produces pre-heated water under pressure for heat energy conservation. Operation of this device is based on simultaneous and/or sequential action of an inductor and an immersion heater and can be easily connected to the electricity network as a single or a three phase unit. Heat energy accumulated by the Heat Production and Storage Device can be utilized in different industrial technological processes during periods of high energy prices.

Self-stability analysis of MHTGRs: A shifted-ectropy based approach

Abstract Because of the strong inherent safety, the modular high temperature gas-cooled nuclear reactor (MHTGR) has been seen as the chosen technology for the next generation of nuclear power plants (NPPs). Self-stability of a nuclear reactor, which is the ability that the reactor state can converge to an equilibrium point without control input, has great meaning in designing control and operation strategies for the NPPs based on MHTGR technology. In this paper, self-stability of the MHTGR is analyzed from a physical viewpoint. A shifted-ectropy method for analyzing the stability of the equilibriums of general thermodynamic systems is firstly established. Based upon this approach, it is proved theoretically that the equilibriums of the MHTGR dynamics are globally asymptotically stable. Numerical simulation results, which illustrate the MHTGR self-stability feature directly, are consistent with the theoretical result.

A methodology to a DB-MHR fuel recharge evaluation—A basic comparison between WIMSD-5B and MCNPX codes

The deep burn modular helium reactor (DB-MHR) belongs to the new generation of nuclear power plants. Such reactor presents the possibility to use reprocessed fuels and a special refueled based on shuffling scheme that displaces the fuel pins after each cycle. This work presents a simplified methodology to evaluate qualitatively the shuffling scheme of these reactors using the WIMSD-5B code. The keff behavior during the burnup was verified by the WIMSD-5B and the MCNPX 2.6.0 codes using three types of core geometry. The WIMD-D5 and the MCNP presented very similar results.

Design of condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) of APR+ (Advanced Power Reactor Plus)

The APR+ (Advanced Power Reactor Plus), a next generation nuclear power plant in Korea, has adopted the PAFS (Passive Auxiliary Feedwater System) on the secondary system of the steam generator (SG) as an advanced safety feature. It is intended to replace the conventional auxiliary feedwater system, which consists of active components for the SG in a passive way. It removes decay heat from the reactor core by cooling down the secondary system of the SG using a condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank).The objective of this study is to design a condensation heat exchanger for the PAFS and to evaluate the cooling performance for the proposed design using the thermal hydraulic system analysis code, MARS (Multi-dimensional Analysis for Reactor Safety). Requirements such as the heat removal capacity and the prevention of water hammer were preferentially considered to determine the design parameters of the heat exchanger tube. The MARS code analysis result showed that the proposed design of the PAFS heat exchanger is able to cool down the required amount of decay heat. The distribution of a liquid volume fraction and flow regime predicted by the MARS code shows that the proposed design of the heat exchanger excludes the water hammer inside the tube. Estimation of a two-phase flow pressure drop indicates that the pressure drop inside the tube is negligible compared to the total pressure drop in the PAFS. From the MARS code analysis, it is concluded that the proposed design of the condensation heat exchanger in the PAFS satisfies the overall criteria for the performance of the passive heat removal system in APR+.

Preparation and Thermal Stability of a Mechanically Alloyed Oxide Dispersion Strengthened Ferritic Steels

Oxide dispersion strengthened ferritic steels (so-called nanostructured ferritic alloys, NFAs), which are candidate structural materials in next generation nuclear power plant, have attracted much attention during recent years. In this work, iron oxide as oxygen carrier and titanium, yttrium hydrides were together mechanically milled with Fe-14Cr-3W gas-atomized powder. The thermal stability and recrystallization behaviour of the as-milled ferritic powder were studied by means of metallography, SEM, TEM and microhardness test. After ball milling for 48h, complete solid solution of bcc-Fe was formed in the as-milled powder. The thermal analysis results show that dispersed oxides with an average diameter of 5nm precipitate from the supersaturated matrix at about 850 °C. During annealing at temperatures from 800 to 1000 °C, a large number of equiaxed grains as fine as few hundreds of microns were found embedding in the matrix; the recrystallized grains stay quite stable and show minor dependence on annealing temperature and time. After being heated to 1200 °C for extended time, abnormal grain growth took place and resulted in bimodal grained structure. The effect of secondary particles on the thermal stability and recrystallization behavior of the ferritic steel was also discussed.

A method for techno-economic analysis of supercritical carbon dioxide cycles for new generation nuclear power plants

The energy generation industry faces new challenges and requirements. The fourth generation of nuclear power plants is developed to meet the future clean energy demand by 2030. The supercritical carbon dioxide (S–CO2) Brayton closed cycle features compactness and high efficiency, combined with short construction time, standing for higher profitability when compared to current steam cycles. In this article, a method for assessing overall viability of direct and indirect carbon dioxide cycles is described. This method comprises both thermodynamic and economic analyses. Thermodynamic and economic models have been developed to assess several cycle designs in terms of overall viability that includes thermodynamic performance and financial outcome. Each cycle has been calculated through an iterative process that takes into account the variation of the properties of the working fluid and returns temperatures, pressure, efficiency, and power generation for the given design characteristics. The economic model, on the other hand, includes the calculation of overnight cost, which is a function of the cycle choice, through the cost of the components. Furthermore, plant profitability is calculated over a given period, which depends on both overnight cost and power plant efficiency. The method has been applied on an indicative parametric analysis of two S–CO2 cycles (direct and indirect) where the capability of identifying optimum areas has been case shown. In terms of overall pressure ratio, it was found that a thermodynamically optimized cycle would not coincide with an economically optimized one, due to the effect of component costing. On the other hand, turbine entry temperature appeared to considerably improve both power output and economic outcome, even though a maximum value is dictated by thermal and mechanical resistances of materials. In overall, even though power generation train is only a subsystem in a nuclear plant, it has been demonstrated that thermodynamic cycle design can have a significant impact on maximizing the earning capacity of the plant, especially through the implementation of a novel highly dense fluid, such as the S–CO2, which bears the potential of high compactness.

Mesoporous materials in the field of nuclear industry: applications and perspectives

In the context of increasing global energy demand associated with high prices for classical fossil fuels and rising concerns about CO2 emissions, nuclear power is attracting renewed interest. Despite the recent accident in the Fukushima power plant, numerous countries maintain very ambitious nuclear programmes. The Generation IV International Forum [A Technology Roadmap for Generation IV Nuclear Energy Systems, Generation IV International Forum, 2002] (an initiative launched by the US Department of Energy in 2000 to coordinate and promote efforts to design the next generation of nuclear power plants) supports works to develop new power plant designs and innovative concepts in the nuclear fuel cycle. These new concepts aim at achieving higher levels of security and resource efficiency as well as a higher resistance towards nuclear proliferation. Because of this blossoming of new ideas, mesoporous materials, which are already widely found in applications such as catalysis, sensing and optics, are expected to make their way in the field of nuclear power production. Up to now, some (hybrid) mesoporous solids have already been investigated as solid ligands to remove actinides and fission product contaminants from liquid effluents, but also as model materials to investigate radiation defects, as possible nuclear waste disposal form and as functional materials to be placed in or close to new forms of nuclear waste matrices. This contribution aims at reviewing the applications of mesoporous materials already described in the field of nuclear industry, and to underline some promising research directions in this area.

Prompt Gamma Emission in Resonance Neutron Induced Fission of 239Pu

Abstract The scientific interest in the resonance neutron induced capture and fission reactions on 239 Pu is continuously rising during the last decade. From a practical point of view, this is because more precise data on capture and fission cross sections, fission fragment mass and kinetic energy distributions, variation of prompt fission neutron and gamma yields in the resonance neutron region, are needed for the modelling of new generation nuclear power plants and for nuclear spent fuel and waste transmutation. From a heuristic and fundamental point of view, such a research improves our knowledge and understanding of the fission phenomena itself. To achieve these goals more powerful neutron sources and more precise fission product detectors have to be used. At the Joint Institute for Nuclear Research (JINR) Frank Laboratory of Neutron Physics (FLNP), where already half a century the thermal and resonance neutron induced nuclear reactions are studied, a new electron accelerator driven white spectrum pulsed neutron source IREN has been built and successfully tested. The improved characteristics of this facility, in comparison with those of the former pulse neutron fast reactor IBR-30, will allow measuring some of the neutron-nuclear reaction data with better precision and accuracy. A new experimental setup for detecting gamma rays (and neutrons) has been designed and is under construction. It will consist of 2 rings (arrays) of 12 NaI(Tl) detectors each (or 1 array of 24 detectors) with variable ring diameter and distance between both rings. Such a setup will make possible not only to measure the multiplicity, energy and angular anisotropy of prompt fission gammas, but also to separate the contribution of prompt fission neutrons by their longer time-of-flight from the fissile target to the detectors. The signals from all the 24 detectors will be recorded simultaneously in digitized form and will be stored on the hard disk of the personal computer for further off-line analysis. The measurement of the prompt gamma-ray emission from 239 Pu resonance neutron induced fission is one of the most probable candidates for the first experiments to be performed at IREN using the newly designed gamma-ray detector.