控制棒内中子注量率精细分布测量

我们采用小型固体核径迹探测器实现了多测点布置及有效地消除了缝隙泄漏中子影响等技术关键。成功地测得了小直径控制棒内中子注量率精细分布。 1.实验原理和方法 把由~(235)U电镀靶片和天然白云母片组成的探测器置入被测样品内。在反应堆中辐照后云母片上形成的径迹密度与所处位置的中子注量率成正比关系,     

Ag-In-Cd芯体辐照后化学成分的计算

Ag-In-Cd芯体的辐照肿胀规律是评价控制棒堆内运行安全性的基础。辐照后Ag-In-Cd芯体的成分变化是引起肿胀的主要物理原因,但目前尚未看到定量计算成分变化的研究报道。本文根据Ag-InCd芯体中各核素的嬗变反应方式和反应截面,建立描述芯体成分变化的微分方程组,编写该微分方程组的数值计算程序,计算预测芯体成分随热中子注量的变化规律。当热中子注量为1.5×1021~6.2×1021 cm-2时,芯体中各元素的含量与热中子注量之间呈较好的线性关系,而芯体表层Sn和Cd的含量会达到中心含量的2倍以上。     

Ag-In-Cd合金辐照后的微观组织变化

Ag-In-Cd合金在核电站压水堆控制棒中广泛使用,其辐照肿胀行为是评价Ag-In-Cd控制棒使用寿命的关键因素。本文通过制备不同成分的模拟合金,来模拟Ag-In-Cd合金在堆内辐照后的成分变化,分析合金的密度及微观组织特点。结果发现,当Ag含量低至77.5%(质量分数)时,合金会分解为fcc和hcp两相,fcc相中贫Sn高Ag,hcp相中富Sn低Ag。当Ag含量在55%~61%之间时,合金以hcp单相存在。由实测的密度拟合出了合金密度随成分变化的关系式。此结果对于理解和掌握Ag-InCd合金的辐照肿胀行为有重要意义。     

可逆比热容法测量反应堆中子注量率的可行性及应用前景

热分析仪器和测量技术的迅速发展为通过测量受辐照材料热性质的变化测量中子注量提供了可能。本文提出采用调制差示扫描量热（MDSC）法测量反应堆辐照的含硼材料可逆比热容的变化,进而得到反应堆的中子注量率。从理论和实验两方面讨论了利用该方法测量反应堆中子注量率的可行性。介绍了可逆比热容法测量反应堆中子注量率的原理和实验方法。展望了这种测量方法在测量高注量反应堆中子注量率的应用前景。     

能谱变化对活化法中子注量测量的影响

在反应堆中子注量测量中,活化探测器可能会经历多个燃料循环的中子辐照,不同燃料循环的中子能谱也会发生变化。考虑到中子能谱变化的影响,对某批次国产反应堆压力容器辐照材料进行中子注量测量修正。计算结果表明,探测器权重快中子注量率（E＞1.0 MeV）修正后比理论中子注量率（E＞1.0 MeV）高1.75%;与修正前相比降低了3.73%,中子能谱变化的影响不容忽视。      

HFETR占栅元铍中孔控制棒物理特性研究

研究了高通量工程试验堆（HFETR）占栅元铍中孔控制棒物理特性。首先,采用CELL程序计算各组件的少群截面参数;然后分别对占栅元控制棒和占栅元铍中孔控制棒进行了堆芯物理计算,并对反应堆轴向热中子注量率分布、60Co产量以及控制棒价值做了比较。研究结果表明,占栅元铍中孔控制棒完全可以用于HFETR的反应性控制,而且可以提高反应堆的安全性和经济性。      

基于人工神经网络的RPV材料辐照脆化预测模型研究

在分析一定量随站测试样品的基础上,构建了具有较高精度的反应堆压力容器（RPV）材料韧脆转变温度（DBTT）预测的人工神经网络模型,并利用模型研究了中子注量和中子注量率对RPV材料DBTT的影响。结果表明,材料DBTT随着中子注量增加出现先线性上升,然后平缓上升,最后饱和的趋势,而中子注量率对RPV材料辐照脆化的影响不明显。      

可逆比热容法测量高中子注量率

目前直接测量高中子注量率一直难于实现，为了解决该问题，本文采用热分析仪器测量受中子辐照后材料可逆比热容的变化来测量高中子注量率。对中国先进研究堆(CARR)辐照孔道的高中子注量率进行了测量，所测量的中子注量率与参考值均在3%左右符合，证明可逆比热容法可直接测量高中子注量率。本文方法是对高中子注量率直接测量的有益尝试。     

用固体径迹探测器验证公式Φ_U（r）／_U≈Φ_M（r）／_M

利用固体径迹探测器测量反应堆不同位置燃料元件内的中子注量率，得到反应堆燃料元件内的中子注量率分布。与对应点慢化剂内中子注量率进行比较，对反应堆物理实验中一个近似假设公式ΦＵ（ｒ）／ΦＵ≈ΦＭ（ｒ）／ΦＭ进行了验证。给出了该公式成立的条件。     

TIMS测定辐照后Gd同位素比的方法研究

建立了热电离质谱法(TIMS)测量天然及辐照后氧化钆同位素丰度比的检测方法。天然氧化钆制备成靶件,放入高通量工程试验堆(HFETR)预定孔道接受中子辐照,辐照时间共计91.3 h,反应堆功率为80 MW,辐照孔道中子注量率约为2×10~(14) n·cm~(-2)·s~(-1)。辐照后靶件经切割、转运、溶解、制样与测量等过程,完成了对可燃中子毒物钆的辐照与测量,并对测量值进行了修正。数据结果表明,热电离质谱法对Gd辐照后检验的分析数据准确可靠,辐照后样品的后处理方法合理,钆的各同位素的丰度变化值前后吻合,且与中子吸收截面大小密切相关。本方法可用于铀钆混合燃料芯块和卸料元件组件中可燃毒物钆的同位素分析。获得的实测数据可反向用于理论计算修正,以期获得更优的反应堆堆芯设计方案。     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

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.     

超级均匀化方法用于球床氟盐冷却高温堆扩散计算

球床氟盐冷却高温堆的控制棒位于侧反应射层内,存在无裂变中子源且受堆芯泄漏谱强烈影响的强吸收体区域扩散计算难题。超级均匀化方法(Super Homogenization,SPH)被用于对氟盐球冷却床堆侧反射层中控制棒区域的强吸收体进行等效均匀化处理,同时堆芯除控制棒区域外采用谱修正方法(Spectra Modification,SM),将输运计算的结果作为基准进行验算。结果表明,SM-SPH模型能有效地计算球床氟盐冷却高温堆反射层控制棒价值及通量分布,并且较常规的SPH方法能更好地处理棒间干涉效应。     

Existence of the fifth unstable nuclided series

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Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity (ρ), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.     

启明星Ⅱ号中子吸收体反应性价值测量

为验证加速器驱动的次临界系统（ADS）次临界反应堆设计时理论计算所使用的计算程序和核数据,在ADS启明星Ⅱ号零功率装置的铅冷堆芯中采用不锈钢元件作为中子吸收体,利用周期法对不锈钢中子吸收体的反应性价值进行实验研究。实验结果表明：吸收体的反应性价值随元件与中心径间距离的增加而降低,实验测量与理论计算的反应性价值接近,变化趋势相互吻合。经实验验证的理论计算程序和核数据可用于ADS次临界反应堆的设计。     

Effect of <Superscript>3</Superscript>He and <Superscript>6</Superscript>Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of ³He and ⁶Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating ³H, ³He, and ⁶Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The ³He and ⁶Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 84–88, February, 2008.     

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B₄C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH_1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.     

Neutron fluctuations induced by absorber vibrations: the QMC reactor experiment—I

The dependence of rod vibration induced, neutron density fluctuations on the static neutron gradients was investigated experimentally at the QMC research reactor. By appropriate fuel loading arrangements the rod was made to vibrate (a) within a flat flux and (b) a flux gradient. The neutron density was fluctuating with the frequency of the vibrating absorber only when the static flux gradient in the vicinity of the absorber was not zero. The double frequency effect was not observed.