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The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained. 相似文献
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Analytical assessments, associated with the choice of the unit capacity of a serially built fast reactor under conditions
of the future advancement of nuclear power, are presented. It is shown that considering the limited resources of natural uranium,
the development of a reliable raw materials base must be based on the development of fast reactors with expanded breeding
of fuel and fuel cycle closure. Since fast reactors, together with energy production, are also producers of new fuel, their
parameters must be optimized taking account of this factor on the basis of systems analysis. Calculations show that the optimal
capacity for fast reactors is in the 1 GW range.
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Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 83–88, August, 2007. 相似文献
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熔盐快堆增殖是当前国际上关注的热点,本文基于堆芯结构双流体方案,利用氟化或氯化熔盐中铀钚重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。对铀钚燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+PuF_4+UF_4、NaF+PuF_4+UF_4和NaCl+PuCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数。分析了增殖比BR(Breeding Ratio)受反应堆裂变区、增殖区和中子反射层的尺寸影响,熔盐中~6Li和~(35)Cl同位素丰度对BR的影响,以及BR随运行时间动态变化。计算结果表明:氯盐方案(BR=1.46)与两种氟盐方案(BR≈1.06)相比较,具有更大的增殖能力优势。结合熔盐相图、BR随重金属摩尔浓度变化和BR最大值随熔盐平均工作温度变化曲线,可以在熔盐快堆设计中快速确定熔盐的工作温度、重金属摩尔浓度和反应堆增殖比。 相似文献
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熔盐快堆是当前国际上关注的热点之一,本文基于堆芯结构双流体方案,即裂变熔盐燃料和增殖熔盐介质各自独立冷却循环,利用氟化或氯化熔盐中钍铀重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。通过比较钍铀燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+ThF_4+UF_4、NaF+ThF_4+UF_4和NaCl+ThCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数,分析了增殖比BR(breeding ratio)受反应堆裂变区、增殖区和ZrC中子反射层的尺寸影响、熔盐中~6Li和~(35)Cl同位素丰度的影响,以及熔盐密度误差对BR计算值的准确性影响、易裂变核素随反应堆运行时间演化等。在钍铀燃料循环熔盐快堆中,通过优化处理得到三种熔盐燃料方案的增殖比BR约为1.2。 相似文献
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小型模块化熔盐快堆燃料管理初步分析 总被引:1,自引:0,他引:1
由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。 相似文献
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A way of development to standardize a small fast nuclear reactor system, which is considered one of the suitable concepts at next generation for satisfying such needs as generality, small dependence on natural resources, safety and non-proliferation, is proposed. This process consists of three steps : the first is to demonstrate the basic system within a short period based on current techniques, the second is to achieve greatly higher economy, and the final is to standardize the commercial system that can economically compete with or overcome current light water reactors. A technical investigation is conducted on the performance of a mixed-oxide (MOX)-fueled small fast reactor with a reflector-driven reactivity control system to satisfy the needs at the first step, considering plenty of accomplishments on the MOX fuel and its advantage for limiting the duration of development to the level required at the stage. The results obtained from a series of neutronic and thermal-hydraulic calculations show the feasibility of a small fast reactor that produces the electric power of about 50MW, achieves about two-year consecutive operation with high safety performance and is greatly flexible for updating the system. A mixed-nitride-fueled core is found to be promising past the first stage. 相似文献
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Liao-Yuan He Yong Cui Liang Chen Shao-Peng Xia Lin-Yi Hu Yang Zou Rui Yan 《核技术(英文版)》2023,(3):156-172
Due to their unique features, such as the inherent safety, simplified fuel cycle, and continuous on-line reprocessing, molten salt reactors (MSRs) are regarded as one of the six reference reactors in the Generation IV International Forum (GEN-IV).Molten chloride salt fast reactors (MCFRs) are a type of MSR. Compared to molten fluoride salt reactors (MFSRs), MCFRs have a higher solubility of heavy metal atoms, a harder neutron spectrum, lower accumulation of fission products (FPs), and better bre... 相似文献
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P. V. Andreev G. M. Gryaznov E. E. Zhabotinskii A. M. Nikonov V. I. Serbin 《Atomic Energy》1991,70(4):275-279
Conclusion The use of a thermionic NPS with a thermal reactor in space technology to supply power to the RMPS offers broad possibilities for interorbital delivery of payloads while using comparatively cheap launch rockets to place spacecraft in a fixed orbit. The flight time from a fixed to a geostationary orbit ranges from several months to half a year, and the mass of the payload in a geostationary orbit for optimal RMPS parameters may reach 7–8 tons (not counting the mass of the NPS).It should be noted that after the flight is completed, the NPS can serve as a source of electrical power for spacecraft in geostationary orbit.Red Star Scientific-Production Organization. Translated from Atomnaya Énergiya, Vol. 70, No. 4, pp. 221–224, April, 1991. 相似文献
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Douglas E. Burkes Randall S. Fielding Douglas L. Porter 《Journal of Nuclear Materials》2009,392(2):158-163
Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt. 相似文献
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Thermal fatigue crack growth in a fast breeder reactor is theoretically investigated with the aid of probabilistic fracture mechanics (PFM) under the conditions that (i) the temperature variation is a narrow-band stationary process and (ii) the crack grows owing only to the peak stress variation. First, a statistical property of residual life of the component with single crack is derived in an analytical form with the aid of an extended Markov approximation method, which is an efficient mathematical technique in PFM. Next, discussion is carried out on the generalization of the primitive model to the case with plural cracks, where a stress relaxation factor is introduced to express a stress intensity factor of each crack. Finally, a numerical example is shown to examine the quantitative behavior of the component's residual life, and sensitivity analysis is performed with respect to some model parameters. 相似文献
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P.V. Girija Shankar 《Nuclear Engineering and Design》1977,44(2):269-277
In applying optimal control theory to a boiling water nuclear reactor (BWR) system which includes the primary recirculation loop, the turbine and their associated auxiliaries, it is necessary to have a linearized mathematical model. Nonlinear and linearized models of a turbine coupled to a BWR, and open-loop responses for specific disturbances are presented. 相似文献
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V. I. Matveev V. M. Murogov A. V. Zhukov E. S. Egorova A. S. Seregin M. Izbashesku M. Dobrean G. Gavrush M. Poshirka 《Atomic Energy》1991,70(6):504-506
Power-Physics Institute. Romanian Institute of Nuclear Power, Pitesh, Romania. Translated from Atomnaya Énergiya, Vol. 70, No. 6, pp. 403–405, June, 1991. 相似文献