The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:
• A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.
• Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a ‘reference design’, developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the ‘reference design’ was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to ‘calibrate’ the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly.
• Preliminary selection was made for the HPLWR scale, boundary conditions, core and fuel assembly design, reactor pressure vessel, containment, turbine and balance of plant.
• A review of potentially applicable materials for the HPLWR was completed and a preliminary selection of potential in-vessel and ex-vessel candidate materials was made.
• A thorough review of heat transfer at supercritical pressures was completed together with a thermal-hydraulics analysis of potential HPLWR sub-channels. This analytical tool supports the core and fuel assembly design.
• The RELAP5 and the CATHARE 2 codes are being upgraded to supercritical pressures. Thus they can be used to support the HPLWR core design and to perform plant safety analyses.
• Assessment of the HPLWR design constraints, based on current LWR technology was documented. This document stresses the various criteria that must be satisfied in the design (e.g. material, temperature, power, safety criteria, etc.) based on experience gained in the design of PWR.
• Preliminary economic assessment concluded that the HPLWR has the potential to be economically competitive. However, an accurate assessment can only be done after the HPLWR design has been fixed. A more accurate economic assessment may be performed after the conclusion of this project.
Article Outline
1. Introduction
2. Approach
3. Project objectives and work program
4. Main achievements
4.1. Heat transfer at supercritical pressures
4.2. Cladding materials
4.3. Core design
4.4. Reactor pressure vessel design
4.5. Overall plant concept
4.6. Transient safety analyses
5. Conclusions
Acknowledgements
References
1. Introduction
In view of continuous industrial expansion in industrially developed countries, the need to accelerate the development of underdeveloped countries, the deregulation of electric utilities, the desire to reduce global warming (believed to be directly related to the amount of CO2 in the atmosphere and therefore to be caused by combustion of fossil fuel)—there is a renewed prospect that nuclear energy will once again be in demand. Evidence of this trend is already seen in the United States. During the last 3 years, the US Department of Energy (US DOE) has led an ambitious program, named Generation IV nuclear reactors, with the main objective to help revitalize the nuclear energy option. In order for nuclear energy to be a viable economical option, there is also a continuing need to improve the economics and efficiency of light water reactors (LWR) similarly to the improvements made in fossil power plants.The concept in this HPLWR project, as described by Heusener et al., 2000a and Heusener et al., 2000b, involves an LWR operating in thermodynamically supercritical regime. In a once-through thermodynamic cycle, the water enters the reactor as water and exits as high-pressure steam without change of phase. Consequently, it is expected that this may lead to a simplified plant design. The concept of the once-through supercritical-pressure light water cooled reactor has been studied by the University of Tokyo over the past decade, as reported by Oka and Koshizuka, 1996 and Oka and Koshizuka, 1998, Dobashi et al. (1998), and Lee et al. (1999). It has been reviewed by the Tokyo Electric Power Company and other Japanese industrial companies, and reported by Tanaka et al. (1997). The main advantages of a reactor cooled and moderated by supercritical water are that above the critical pressure of water (22.1 MPa) supercritical water does not exhibit a change of phase and the heat is effectively removed at or above the pseudo-critical temperature that corresponds to the boiling point at sub-critical pressure (385 °C at 25 MPa). Thus, steam-water separation is not necessary at the core exit and the turbines can be driven directly by the high temperature coolant leaving the core.On the other hand, the development of this reactor concept has to account for the high temperatures that are expected to be achieved as well as for the large axial density gradient within the core. Therefore additional research and development effort is expected to be devoted in particular to these areas.As an example of such a system operating at 25 MPa, the coolant enters the core at 280 °C. It exceeds the pseudo-critical temperature as it flows upward through the core and it exits the core at 508 °C directly into the steam turbines. The thermal efficiency of such a cycle is approximately 44% and is strongly affected by the core outlet temperature (see Fig. 1). The development effort carried out by Dobashi et al. (1998) has been primarily conceptual in nature and has pointed out the potential merit of the once-through concept. By using these results as a starting point, it is possible to reach a conclusion on whether or not the once-through supercritical-pressure LWR is economically and physically a viable solution that may help sustain the nuclear option. Because of the expected high efficiency, high-temperature, high-pressure and high power density we have named this concept high performance light water reactor (HPLWR). 相似文献
Experimental work on the ternary system Ti-Hf-B as well as on the two binary boundary systems Hf-B and Ti-Hf is reviewed.
The individual phases of these systems are described with thermodynamic models. The models were fitted to the experimental
information and the resulting optimized data sets are presented. The calculated thermodynamic properties of the two binary
compounds HfB and HfB2 are compared with experimental measurements. Phase diagrams were calculated for the bi-nary systems Hf-B and Ti-Hf as well
as for the ternary Ti-Hf-B system; for the latter, several isother-mal sections, isopleths, and the liquidus surface are included.
They are presented and compared with the experimental information. A reaction scheme for the ternary Ti-Hf-B system was constructed. 相似文献
For future nuclear power plants possible core melt scenarios have to be taken into account in the plant layout. Different solutions for core melt retention are proposed in the literature. The core melt retention devices should as far as possible be simple in construction, to minimize the deepness of sophistication of the physical problems connected with it and to minimize unforeseen technological problems. Furthermore it should be in good compliance with normal operational needs.One very promising solution seems to be a large enough spreading of the melt with water cooling from above. This is the favoured solution for the development of the French-German European pressurized water reactor. The technical concept and the problems to be expected in the course of an anticipated core melt accident (to be overcome in an actual design) are shown. The needs for analytical and experimental support in the early conceptual phase and later on in the design stage of plant construction are discussed. Typical analytical and first experimental results for the large core melt spreading are presented. 相似文献
An open-cell metallic foam was employed as an analogue material for human trabecular bone to interface with polymethyl methacrylate (PMMA) bone cement to produce composite foam–cement interface specimens. The stress-displacement curves of the specimens were obtained experimentally under tension, shear, mixed tension and shear (mixed-mode), and step-wise compression loadings. In addition, under step-wise compression, an image-guided failure assessment (IGFA) was used to monitor the evolution of micro-damage of the interface. Microcomputed tomography (µCT) images were used to build a subject-specific model, which was then used to perform finite element (FE) analysis under tension, shear and compression. For tension–shear loading conditions, the strengths of the interface specimens were found to increase with the increase of the loading angle reaching the maximum under shear loading condition, and the results compare reasonably well with those from bone–cement interface. Under compression, however, the mechanical strength measured from the foam–cement interface is much lower than that from bone–cement interface. Furthermore, load transfer between the foam and the cement appears to be poor under both tension and compression, hence the use of the foam should be discouraged as a bone analogue material for cement fixation studies in joint replacements. 相似文献
A simple control system for induction motor drives with the current source inverter is presented. This system is based on the idea of field-orientated control. In contrast to well-known ways of realization a torque proportional quantity is calculated from stator voltage and current without explicit flux calculation or integration. A speed sensor is not required. The theoretical principles of this method are discussed. Some transients which illustrate the performance of some different system variants are presented. 相似文献
Based on an assessment of the available experimental thermochemical and phase diagram information available, the phase equilibria of the C-Hf-Zr system were calculated. The G of the individual phases was described with thermodynamic models. The liquid phase was described as a substitutional solution using the Redlich-Kister formalism for excess G. Graphite was treated as a stoichiometric phase. The solid solutions of carbon in α(Hf,Zr) and β(Hf,Zr), as well as the non-stoichiometric phase (Hf,Zr)C1?x, were represented as interstitial solid solutions using the compound energy model with two sublattices. The parameters in the models were determined by computerized optimization using selected experimental data. A detailed comparison was made between calculation and experimental data. 相似文献
The increase of steam parameters to supercritical conditions could reduce the power generating costs of light water reactors significantly [Proceedings of SCR-2000 (2000) 1]. Core assemblies, however, will differ from current BWR or PWR design. In this context, this paper summarizes the main results related to a thermal-hydraulic design analysis of applicable fuel assemblies. Starting from a thorough literature survey on heat transfer of supercritical fluids, the current status indicates a large deficiency in the prediction of the heat transfer coefficient under reactor prototypical conditions. For the thermal-hydraulic design of such fuel assemblies the sub-channel analysis code Sub-channel Thermal-hydraulic Analysis in Fuel Assemblies under Supercritical conditions (STAFAS) has been developed, which will have a higher numerical efficiency compared to the conventional sub-channel analysis codes. The effect of several design parameters on the thermal-hydraulic behaviour in sub-channels has been investigated. Based on the results achieved so far, two fuel assembly configurations are recommended for further design analysis, i.e. a tight square lattice and a semi-tight hexagonal lattice. 相似文献