Tài liệu miễn phí Năng lượng
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Modeling of hydrodynamic processes at a large leak of water into sodium in the fast reactor coolant circuit
In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodiumewater steam generators in fast neutron reactors.
7/4/2019 6:04:47 AM +00:00
Mode control of guided wave in magnetic hollow cylinder using electromagnetic acoustic transducer array
The aim of this work is to demonstrate a method for exciting and receiving torsional and longitudinal mode guided waves with an electromagnetic acoustic transducer (EMAT) ring array.
7/4/2019 6:04:33 AM +00:00
Modal characteristic analysis of the apr1400 nuclear reactor internals for seismic analysis
Reactor internals are sensitive to dynamic loads such as earthquakes and flow induced vibration. Thus, it is essential to identify the dynamic characteristics to evaluate the seismic integrity of the structures. However, a full-sized system is too large to perform modal experiments, making it difficult to extract data on its modal characteristics. In this research, we constructed a finite element model of the APR1400 reactor internals to identify their modal characteristics.
7/4/2019 6:04:12 AM +00:00
Microstructure and mechanical strength of surface ods treated zircaloy 4 sheet using laser beam scanning
The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and Y2O3 particles of 10 μm were selected for ODS treatment using LBS.
7/4/2019 6:03:56 AM +00:00
Method for the analysis of temporal change of physical structure in the instrumentation and control life cycle
The design of computer-based instrumentation and control (I&C) systems is determined by the allocation of I&C functions to I&C systems and components. Due to the characteristics of computer-based technology, component failures can negatively affect several I&C functions, so that the reliability proof of the I&C systems requires the accomplishment of I&C system design analyses throughout the I&C life-cycle.
7/4/2019 6:03:38 AM +00:00
Metal fuel development and verification for prototype generation IV sodium-cooled fast reactor
Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. UeZr fuel is a driver for the initial core of the PGSFR, and U etransuranics (TRU)eZr fuel will gradually replace UeZr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of UeZr fuel, work on UeZr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests.
7/4/2019 6:03:15 AM +00:00
Mechanical properties of two way different configurations of prestressed concrete members subjected to axial loading
In order to analyze the mechanical properties of two-way different configurations of prestressed concrete members subjected to axial loading, a finite element model based on the nuclear power plant containments is demonstrated. This model takes into account the influences of different principal stress directions, the uniaxial or biaxial loading, and biaxial loading ratio.
7/4/2019 6:02:59 AM +00:00
Measuring situation awareness of operating team in different main control room environments of nuclear power plants
In this regard, the operating team's SA in a conventional and digital MCR should be measured in order to assess whether the new design features implemented in a digital MCR affect this parameter. This paper explains the team SA measurement method used in this study and the results of applying this measurement method to operating teams in different MCR environments. The paper also discusses several empirical lessons learned from the results.
7/4/2019 6:02:41 AM +00:00
Mass transfer experiments for the heat load during in vessel retention of core melt
The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.
7/4/2019 6:02:26 AM +00:00
Managing a prolonged station blackout condition in AHWR by passive means
The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.
7/4/2019 6:02:09 AM +00:00
Linear instability analysis of a water sheet trailing from a wet spacer grid in a rod bundle
The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.
7/4/2019 6:01:49 AM +00:00
Leak-before-break analysis of thermally aged nuclear pipe under different bending moments
The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered.
7/4/2019 6:01:32 AM +00:00
Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario
This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012e2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions.
7/4/2019 6:01:15 AM +00:00
KTM tokamak operation scenarios software infrastructure
The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.
7/4/2019 6:00:56 AM +00:00
Korean students’ behavioral change toward nuclear power generation through education
This indicates that education can be effective in promoting widespread social acceptance of nuclear power and its continued use. In order to induce behavior change toward positive judgments on nuclear power generation, it is necessary to focus on attitude improvement while providing the information in all areas related to the perception, knowledge, attitude, and behavior. Here, the positive message on the convenience and the safety of nuclear power generation should be highlighted.
7/4/2019 6:00:39 AM +00:00
Key findings from the artist project on aerosol retention in a dry steam generator
To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.
7/4/2019 6:00:23 AM +00:00
Irradiation performance of U-Mo monolithic fuel
In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.
7/4/2019 6:00:00 AM +00:00
Irradiation device for irradiation testing of coated particle fuel at HANARO
The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle) during irradiation testing.
7/4/2019 5:59:42 AM +00:00
Investigation of characteristics of passive heat removal system based on the assembled heat transfer tube
The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.
7/4/2019 5:59:24 AM +00:00
Investigation of burst pressures in pwr primary pressure boundary components
In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization.
7/4/2019 5:59:02 AM +00:00
Investigation of a hydrogen mitigation system during large break loss of coolant accident for a two-loop pressurized water reactor
The management of hydrogen safety and prevention of overpressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity.
7/4/2019 5:58:48 AM +00:00
Investigation of activated carbon adsorbent electrode for electrosorption based uranium extraction from seawater
The objective of this research is to investigate the use of new adsorbent material for cost effective uranium extraction from seawater. An activated carbon-based adsorbent material is developed and tested through an electrosorption technique in this research.
7/4/2019 5:58:31 AM +00:00
Investigating dynamic parameters in HWZPR based on the experimental and calculated results
The measured a is successively used to determine the experimental value of the effective delayed neutron fraction as well. According to the experimental results, beff of the HWZPR reactor under study is equal to 7.84e3. This value is finally used to validate the calculation of the effective delayed neutron fraction by the Monte Carlo methods that are discussed in the document.
7/4/2019 5:58:15 AM +00:00
interparticle potential of 10 nanometer titanium nanoparticles in liquid sodium: theoretical approach
The ab initio calculation reveals that a strong repulsive force driven by the solvation potential exceeds the interparticle attraction and predicts the agglomeration energy required for two 10-nm Ti NPs to be 4 1017 J. The collision theory suggests that Ti NPs can be effective suppressors of the SWR due to the high energy barrier that prevents significant agglomeration of Ti NPs in quiescent liquid Na.
7/4/2019 5:57:59 AM +00:00
Interaction studies of ceramic vacuum plasma spraying for the melting crucible materials
Thermal cycling tests showed that the adhesion of the TiC, ZrC, and ZrO2 coating layers with niobium was relatively weak compared to the TaC and Y2O3 coatings. The TaC and Y2O3 coatings had better cycling characteristics with no interconnected cracks. In the interaction studies, ZrC and ZrO2 coated rods showed significant degradations after exposure to U-10 wt.% Zr melt at 1600ºC for 15 min., but TaC, TiC, and Y2O3 coatings showed good compatibility with U-Zr melt.
7/4/2019 5:57:42 AM +00:00
Integrity analysis of an upper guide structure flange
The mode superposition and full transient methods were used to perform timeehistory analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.
7/4/2019 5:57:27 AM +00:00
Identification of nonregular indication according to change of grain size/surface geometry in nuclear power plant (NPP) reactor vessel (RV)-upper head alloy 690 penetration
In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed.
7/4/2019 5:07:59 AM +00:00
An autonomous control framework for advanced reactors
The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision.
7/4/2019 5:07:37 AM +00:00
Acceleration of step and linear discontinuous schemes for the method of characteristics in DRAGON5
The applicability of the algebraic collapsing acceleration (ACA) technique to the method of characteristics (MOC) in cases with scattering anisotropy and/or linear sources was investigated. Previously, the ACA was proven successful in cases with isotropic scattering and uniform (step) sources. A presentation is first made of the MOC implementation, available in the DRAGON5 code.
7/4/2019 5:07:22 AM +00:00
A TakagieSugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation
This paper proposes a TakagieSugeno (TeS) fuzzy logic-based power distribution system. Two TeS fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised TeS fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.
7/4/2019 5:07:04 AM +00:00