共查询到19条相似文献,搜索用时 156 毫秒
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SHB—5临界装置铀水栅堆芯硼微分价值和总后备反应性的测量 总被引:1,自引:1,他引:0
介绍了SHB-5临界装置铀水栅堆芯硼微分价值的测量,给出了利用非线性牛顿迭代法得到的硼微分价值符合曲线和几种典型硼浓度的硼反应性积分价值;同时给出了利用硼微分价值符合曲线得到的控制棒积分价值、可燃毒物棒总价值和堆芯总后备反应性;这些结果与脉冲中子源法测量结果基本符合。 相似文献
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介绍了利用硼中毒法测量微型临界堆芯硼中毒效应的实验研究结果。为了研究硼酸在该堆芯中的中毒效应,利用硼中毒法采用数字反应性仪和自动电位滴定仪及秒表测量了硼微分价值、堆芯临界硼浓度、堆芯后备反应性等。实验结果表明:实测值与理论计算值相符,结果可信;该实验结果可用于验证理论计算,也可为该堆型的带功率运行提供参考。 相似文献
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次临界反应性测量的空间修正及其应用综述 总被引:2,自引:0,他引:2
次临界下的反应性测量技术有着自身的特点,次临界下控制棒的动作、堆芯的次临界度以及外中子源的存在都会对堆芯中子通量的分布产生影响,因此通常情况下堆芯的次临界度只能"监视",无法准确测量。在堆芯模拟软件发展的基础上,国外科研人员提出了次临界下点堆模型的空间修正方法,将这种方法用于动态棒价值测量(DRWM),并在此基础上进一步发展了次临界控制棒价值测量(SRWM),这些技术有的已经被国内核电站使用,但是国内对空间修正的原理及方法鲜有介绍。本文针对这种需求,总结概括了国外商用堆次临界反应性测量的基本原理与方法,并结合反应性测量仪表技术,给出了次临界反应性仪的数据处理流程,这对于推进国内商用堆次临界反应性测量的研究和实际应用具有较为重要的意义。 相似文献
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固定棒位法测量控制棒总价值 总被引:1,自引:1,他引:0
控制棒价值测量的准确度与效率对核电厂的安全性与经济性具有重要影响。在动态刻棒等反应性测量工作中,本底与中子源对探测器有显著影响,致使根据实测电流计算得到的反应性显著偏离真实值。基于点堆逆动态方程,通过对本底与中子源影响的分析,利用固定棒位状态下的测量数据计算反应性并得到控制棒总价值,给出了一种不受本底与中子源影响的简便的控制棒总价值测量计算方法,并在零功率实验装置上进行验证。结果表明,该方法可有效避免本底和中子源组件对反应性探测的影响,并简化了离线理论计算,其与周期法计算结果的相对偏差在1%以内。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(12):1082-1087
The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions. 相似文献
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启明星Ⅱ号铅堆堆芯的首次物理启动旨在完成国内首座铅冷快堆零功率装置的装料与达临界,掌握堆芯安全特性。考虑铅堆堆芯使用两种燃料元件,临界元件数量较大,不同区域的中子能谱与燃料元件价值差异大的特点,首次物理启动对启动中子源与中子计数探测器进行了选取与验证,评价了模拟元件对中子的散射与吸收的影响,制定了分区外推的装料方案。按照装料方案,铅堆堆芯完成了装料,安全实现了首次临界,测量了模拟元件、燃料元件、安全棒和调节棒反应性。本文工作为后续实验运行提供了重要的实验参数与临界装载方案。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(9):662-670
The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (keff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated keff overestimated the experimental data by about 1.0Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor. 相似文献
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启明星Ⅱ号零功率装置(启明星Ⅱ号)所设计的安全控制部件有安全棒和调节棒,这些控制部件是反应堆安全运行的关键。本文采用逆动态反应性计测量的方法对所选定的控制部件的反应性价值进行了实验测量,并与理论计算结果进行了比较。结果表明,安全控制部件的反应性价值的实验测量结果与理论计算结果的相对偏差为4.46%,二者吻合较好。安全棒系统经力学分析评定,结果表明不会出现卡棒现象,能实现快速停闭反应堆的目的。安全棒系统、调节棒系统的机械性能经堆上反复实验验证,各系统性能稳定可靠,重复性好。 相似文献
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Tayyab Mahmood Ishtiaq Hussain Bokhari Masood Iqbal Tariq Mahmood Naseer Ahmed Muhammad Israr 《Progress in Nuclear Energy》2011,53(6):729-735
Neutronic and thermal hydraulic analyses have been carried out for current core of Pakistan Research Reactor-1 (PARR-1). Comparison was made between calculated and measured key neutronic parameters. Reactor core parameters important for reactor operation and safety have been calculated. Calculated neutronic parameters include: excess reactivity, shut down margin, control rod worth, peak power density location, criticality position, peaking factors, neutron flux in fuel elements and neutron flux at irradiation sites in the core. Calculated thermal hydraulic parameters include: steady-state temperatures and peak temperatures at fuel centerline, clad surface and in water coolant. In order to determine safety margins, heat fluxes at Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB) were determined using standard correlations. After assembling the core, performance of the core was also evaluated by experimentation. The core was assembled and some of the core parameters namely: excess reactivity, shut down margin, control rod worth and flux profile at in-core irradiation sites have been measured. On comparison with experimental data, reasonable agreement has been found between the calculated and the measured parameters. 相似文献
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以圆柱形堆芯试验装置(CCTF)为研究对象,采用轻水堆冷却系统事故工况的瞬态行为最佳估算程序(RELAP5)和自主化堆工设计与安全分析程序(LOCUST),开展堆芯功率分布对CCTF C2-SH2(Run54)试验工况再淹没现象影响的评价研究。研究表明:①计算所得下降段压降、堆芯压降、堆芯出口蒸汽质量流量等计算结果与试验结果吻合较好;②对于堆芯1.015 m处平均通道包壳峰值温度的计算,RELAP5和LOCUST程序计算的包壳峰值温度分别为816 K和813 K,试验结果为898 K,计算值比试验值低约82 K,平均通道包壳温度最后稳定在400 K左右,计算结果与试验结果一致。因此,本研究结果表明LOCUST程序能够较好地对大破口失水事故(LBLOCA)中再淹没阶段的瞬态过程进行模拟。 相似文献