共查询到17条相似文献,搜索用时 375 毫秒
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固体径迹探测器测量反应堆功率研究 总被引:2,自引:1,他引:1
在零功率反应堆上利用固体径迹探测器直接测量燃料元件内的裂变率,可得到反应堆的功率。同时测量反应堆某位置的热中子通量密度,继而可得到单位功率的热中子通量密度。因此,通过测量该点的任何热中子通量密度即可得到反应堆的运行功率。该方法可以减少与能谱测量有关的修正工作。由于辐照所需的中子通量密度低、时间短,因此与活化法等相比具有明显的优点。 相似文献
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巴基斯坦微堆内辐照座通量密度与堆功率的测定 总被引:4,自引:4,他引:0
文章叙述了用4πβγ符合方法,通过测量金箔在堆内照射生成的活性,得到巴基斯坦微型反应堆辐照座内的热中子通量密度。并用积分方法求得裂变率,计算出单位功率的热中子通量密度,建立标准点,最后得到巴基斯坦微堆的功率。 相似文献
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杨波 《核电子学与探测技术》2012,32(5):587-589
介绍了一种高灵敏度裂变室的研制,探讨了其设计方案、制造工艺和性能。这种裂变室灵敏区较宽、热中子灵敏度较高、抗γ能力较强,可作为核反应堆堆外中子注量率测量探测器,可在反应堆启动和不同功率运行时给出功率监测的信号。该裂变室通过了LOCA工况试验测试,可用于事故后监测。 相似文献
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启明星1#次临界装置是我国为开展加速器驱动的次临界系统(ADS)研究而建立的国际上第1个具有快-热耦合结构的次临界反应堆实验装置。启明星1#次临界装置在确定的装载下、由不同能量的外中子源作用时,利用MCNP程序分别对装置快中子能谱区、热中子能谱区燃料元件的径向及轴向裂变率分布进行模拟计算,所使用外中子源的中子能量分别为2.5、5、14MeV。计算结果表明:在外中子源源强相同的情况下,源中子能量越高,裂变率越大;在源中子能量相同的情况下,次临界反应堆的轴向裂变率分布为中间高、两端低,径向裂变率分布在快中子能谱区先减小后增大,而热中子能谱区则是先增大后减小,然后,随着接近反射层又逐渐增大。该裂变率分布计算结果为后续实验测量和探测器布置提供了参考。 相似文献
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Artificial neural networks (ANNs) have recently been utilized in the nuclear technology applications since they are fast,
precise and flexible vehicles to modeling, simulation and optimization. This paper presents a new approach based on multilayer
perceptron neural networks (MLPNNs) for the estimation of some important neutronic parameters (net 239Pu production, tritium breeding ratio, cumulative fissile fuel enrichment, and fission rate) of a high power density fusion–fission
(hybrid) reactor using light water reactor (LWR) spent fuel. A comparison of the results obtained by the MLPNNs and those
found by using the code (Scale 4.3) was carried out. The results pointed out that the MLPNNs trained with least mean squares
(LMS) algorithm could provide an accurate computation of the main neutronic parameters for the high power density reactor. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(3):98-105
A new principle is presented for obtaining absolute reactor power by processing the random fluctuation of neutron flux based on the stochastic nature of nuclear reactions. The required combination of instruments to carry out experiments is described, and experimental results obtained in a swimming pool reactor are reported. The power spectral density of the output current of an ion chamber located near the reactor core is determined by reactor kinetic parameters such as delayed neutron yield, life time, ν (mean number of neutrons generated per fission) and counter efficiency as well as by the total number of neutrons in the core, which is a measure of absolute power. Using either logarithmic amplifier or reactivity meter, absolute reactor power can be measured without any information about detector efficiency. This method has such merits as easiness and simplicity in operation, ability to measure absolute power in the range 0.01~100 W where other methods are inapplicable, and negligible effect of changes in core configuration or in detector position. The results of actual reactor experiments with this method proved to agree fairly well with those of absolute measurement by gold foil activation. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):551-556
The fission rate in the core of the Japan Research Reactor 4 (JRR-4) was determined by a method based on radiochemical analysis of 99Mo formed in the U samples irradiated in the reactor core. The contribution of epithermal neutron fission to the total fission rate was evaluated from the Cd ratio for U fission. The contribution was several percent. For comparison, the thermal neutron flux also was measured, by Au-foil activation. The fission rate determined from the U samples agreed well with the Au-foil data, except at positions in the peripheral region of the reactor core. 相似文献
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F. Tom
ek 《Annals of Nuclear Energy》1987,14(12):677-680
In the present paper, relationships are derived which enable, by using an additional measurement, to calibrate the neutron flux density measured at an unknown low reactor power to the power of 1 MW. The measurements can be carried out with activation detectors using well-known methods, see e.g. (ASTM, 1977, 1980; Zijp, 1984). Constant neutron flux density at a power of 1 MW is assumed during the measurements and a so-called “one-point reactor” model is employed. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):1069-1075
The neutron source introduction method was applied to absolute measurements of low reactor power at the Static Experiment Critical Facility STACY. To obtain the effective neutron source intensity more accurately, which is a key parameter for the source introduction method, the neutron source is newly defined as fission neutrons from the first fission reaction caused by neutrons emitted from the external neutron source. To obtain the newly defined effective neutron source intensity, the probability that a neutron from the external neutron source causes a fission reaction is calculated using the Monte Carlo code MCNP. This calculation took into consideration the three-dimensional complicated core structures. Furthermore, the fission reaction distribution, fundamental mode forward and adjoint flux distribution in a critical state were calculated using the three-dimensional transport code THREEDANT. Following the principle of the neutron source introduction method, an external neutron source was inserted near the STACY core tank and the reactor power was measured. The reactor powers by the neutron source introduction method were in good agreement with the ones from the analyses of the FP activity generated by high power operation. 相似文献
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应用MCNP-4C程序为微型钠冷快堆(MFR)概念设计建立了精确的物理计算模型,并对其临界物理特性、中子注量率分布、功率分布和反应性控制进行详细计算.结果表明:MFR的基本物理特性满足堆芯物理设计要求和设计目标,堆芯功率密度和中子注量率分布均匀合理;控制系统能实现安全有效的反应性平衡,满足反应堆长期运行的需要. 相似文献