共查询到20条相似文献,搜索用时 33 毫秒
1.
《Journal of Nuclear Science and Technology》2013,50(12):1289-1293
A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. 相似文献
2.
3.
控制棒水力驱动系统的设计和研究 总被引:23,自引:2,他引:21
分析了200MW核供热堆控制棒水力驱动系统的设计特点;系统中主要设备的设计特点及特性;旁路自调节结构的设计及其高温下的补偿作用以及系统温度特性的实验结果。经对实验结果的分析表明:HDSCR和各设备的设计合理,运行可靠;各设备的设计不仅降低了设备的加工难度及安装难度,而且改善了系统的温度特性;系统安全能满足200MW核供热堆对控制棒驱动机构的要求。 相似文献
4.
5MW THR控制棒水力驱动系统的设计及实验研究 总被引:1,自引:0,他引:1
控制棒水力驱动系统是不同于一般水动力堆使用的电磁-机械式传动系统的新型传动装置。它以反应堆冷却剂(水)为工作介质,经泵加压后,注入安装在压力壳内的水力步进缸,通过流量来控制水力步进缸外套作步进式运动,拖动与之相联的中子吸收元件。5MW HTR 是世界上首座使用这种传动的反应堆。采用该传动是为得到更好的安全特性,更可靠的驱动特性和良好的经济性。 相似文献
5.
This paper illustrates the work principle of the hydraulic control rod driving system (HCRDS) and the operation of step-up process. As well as the dynamic characteristic of step-up process with the study of experimental results, it also analyzes the influencing parameters of the step-up process, such as holding-flow rate, lift delay time and pressure in front of combined valve, etc. The initial theoretical model of step-up process of HCRDS has been established on the basis of rational analysis, simplification and hypothesis. The calculation result of the theoretical model approximately coincides with the experimental result, which lays a solid foundation for the auxiliary mechanism analysis of HCRDS. 相似文献
6.
控制棒水压驱动系统是清华大学为低温核供热堆NHR200发明的新型的内置式控制棒驱动技术,该驱动系统由水压驱动机构、组合阀、控制棒和缓冲器等组成。控制棒水压驱动系统冷态性能是控制棒步进时间和系统驱动压力选取的基础。本文分析了控制棒水压驱动系统的组成和工作原理,完成了全尺寸控制棒水压驱动系统冷态性能实验,包括水压缸最小落棒压力实验、提升缸带载步进实验和快速落棒实验等。在实验结果的基础上分析了关键特性参数的变化规律和机理。结果表明:最小落棒压力是保持驱动机构销爪正常工作所需的最小驱动压力,其对应于压力时程曲线上峰值波动过程的变化起点;步升和步降过程压力拐点分别对应位移到位点,随着驱动水压的增加,水压缸充压拐点压力逐渐增加,步升时间、充压拐点时间逐渐减少。实验研究成果为控制棒水压驱动系统的设计、优化和工程应用奠定基础。 相似文献
7.
8.
水力驱动控制棒步进动态过程的研究 总被引:1,自引:1,他引:0
通过实验获得了水力驱动控制棒步进的动态过程.揭示了水力驱动控制棒的作用机理,详细分析了其动态特性与控制棒、组合阀性能参数及组合阀操作之间关系.结果表明:控制棒的步进是组合阀输出的流量脉冲、压力波,和步进缸运动产生的大阻尼压力振荡的共同作用过程;步进缸的性能参数限定了其静止平衡状态的流量范围,和其步进过程吸收流量脉冲和压力波的能力,而组合阀的性能参数决定了步进缸静止平衡、延时平衡、流量脉冲和压力波的量值,两者的合理匹配确定了控制棒的步进状态. 相似文献
9.
10.
11.
为提高200MW低温核供热堆经济性,对控制棒结构进行优化设计。在新的控制棒方案中,将控制棒驱动缸移到堆芯活化区以上,控制棒由浮动式活塞带动上下移动。由于驱动缸移出堆芯,燃料组件排布不再缺角,减小了堆的水铀比和堆内的中子吸收,增加了堆的运行时间。适当地加大驱动缸的直径和壁厚,有效降低了制造难度,提高了控制棒运行的可靠性。通过数值计算,分析了上置式水力驱动控制棒的落棒时间。 相似文献
12.
V. F. Kolesov 《Atomic Energy》1964,16(4):377-382
Selected problems concerning the dynamics of a fast pulsed reactor are considered; firstly, the method is discussed of pulse generation in a fast reactor by means of insertion of a rod with the presence in the reactor of a strong neutron source; an equation is derived for the optimum velocity of the rod; solutions are given of dynamical equations relating to the case of movement of the rod; secondly, the effect is investigated of a large internal cavity on the energy release following a pulse and on the reactivity thermal quenching factor for a fast pulsed reactor in spherical geometry.Translated from Atomnaya Énergiya, Vol. 16, No. 4, pp. 309–314, April, 1964 相似文献
13.
Minoru Takahashi Shoji Uchida Yumi Yamada Kazuya Koyama 《Progress in Nuclear Energy》2008,50(2-6):269-275
In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. 相似文献
14.
Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO2 as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10−6 cm3/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed. 相似文献
15.
控制棒水压驱动线是一种新型的内置式控制棒驱动技术,控制棒水力减速装置是水压驱动线的关键部件之一,通过水力减速片和减速筒体的配合对控制棒进行减速,降低快速落棒末端的冲击速度,避免控制棒的变形和损坏。完成了水压驱动线快速落棒减速实验,对减速过程机理进行了分析,在此基础上建立了水压驱动线快速落棒减速理论模型,理论模型的求解结果与实验结果符合很好,从而验证了理论模型的正确性。通过该模型对热态工况下水压驱动线的快速落棒性能进行了分析,为控制棒水压驱动线减速环节的设计和优化奠定了基础。 相似文献
16.
Kyung-Hoon Lee Kang-Seog Kim Jin-Young Cho Jae-Seung Song Jae-Man Noh Chung-Chan Lee 《Nuclear Engineering and Design》2008,238(10):2654-2667
The IAEA's gas-cooled reactor program has coordinated international cooperation for an evaluation of a high temperature gas-cooled reactor's performance, which includes a validation of the physics analysis codes and the performance models for the proposed GT-MHR. This benchmark problem consists of the pin and block calculations and the reactor physics of the control rod worth for the GT-MHR with a weapon grade plutonium fuel. Benchmark analysis has been performed by using the HELIOS/MASTER deterministic code package and the MCNP Monte Carlo code. The deterministic code package adopts a conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation.In order to solve particular modeling issues in GT-MHR, recently developed technologies were utilized and new analysis procedure was devised. Double heterogeneity effect could be covered by using the reactivity-equivalent physical transformation (RPT) method. Strong core–reflector interaction could be resolved by applying an equivalence theory to the generation of the reflector cross sections. In order to accurately handle with very large control rods which are asymmetrically located in a fuel and a reflector block, the surface dependent discontinuity factors (SDFs) were considered in applying an equivalence theory. A new method has been devised to consider SDFs without any modification of the nodal solver in MASTER.All computational results of the HELIOS/MASTER code package were compared with those of MCNP. The multiplication factors of HELIOS for the pin cells are in very good agreement with those of MCNP to within a maximum error of 693 pcm Δρ. The maximum differences of the multiplication factors for the fuel blocks are about 457 pcm Δρ and the control rod worths of HELIOS are consistent with those of MCNP to within a maximum error of 3.09%. On considering a SDF in the core calculations, the maximum differences of the control rod worths are significantly decreased to be 7.7% from 21.5%. It is showed that there are good consistencies between the deterministic code package and the Monte Carlo code from the results of these benchmark calculations. Therefore, the HELIOS/MASTER 2-step procedure can be used as a standard reactor physics analysis tool for a prismatic VHTR. 相似文献
17.
FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:
- 1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.
- 2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.
- 3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.
- 4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.
18.
控制棒水压驱动系统是清华大学为低温核供热堆发明的新型的内置式控制棒驱动技术,控制棒水力减速部件是水压驱动系统的关键部件之一,通过其对控制棒落棒过程进行减速,在保证落棒时间的前提下,降低控制棒快速落棒过程的冲击力。分析了水力减速部件组成和工作原理,确定了水力减速箱侧壁开孔方案,完成了不同开孔方案工况下控制棒水压驱动系统冷态落棒减速性能实验,在实验结果的基础上对比和分析了不同方案下的落棒减速机理和落棒过程特征参数随开孔方案的变化规律。分析结果表明:随开孔面积的增大,落棒时间逐渐减小,落棒峰值速度逐渐增大。在开孔面积大于0.004 m~2时,随开孔面积的增大,落棒峰值速度增大过程趋于平缓,落棒稳定速度和落棒延迟时间变化不大,控制棒触碰碟簧速度缓慢增大。实验研究成果为控制棒水压驱动系统落棒减速部件的理论建模和设计优化提供了基础。 相似文献
19.
M.K. Shoushtari S.M. Sadat Kiai H. Ghaforian 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2010,268(5):519-523
The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems. 相似文献