微型反应堆补偿燃耗的方法

微型反应堆（ＭＮＳＲ）严格限制冷态后备反应性为３．５—４．０ｍｋ，小于０．５β＿（ｅｆｆ）（微堆的β＿（ｅｆｆ）＝０．００８），从根本上杜绝了瞬发临界事故和堆芯元件烧毁事故的发生。在如此小的后备反应性条件下，为了使微堆寿期大于１０ａ，采用间断地添加顶铍反射层的办法来补偿燃耗。理论计算出了添加顶铍反射层厚度与反应性增长的关系，在零功率反应堆上进行了实验校核，并就原型微堆添加顶铍反射层的操作经验作出总结。     

拓宽微堆的应用

针对微堆注量率低 ,运行时间短 ,制备的中短寿命同位素放射性比度低 ,应用困难。从 2 0世纪80年代以来 ,仪器的微量元素分析技术飞速发展 ,造成中子活化分析的市场日益萎缩。深圳大学根据市场的需要 ,不断进行微堆技术改造。将处于微堆侧面铍反射层中的内辐照管改为超热辐照管和添加顶部铍反射层 ,提高了后备反应性。建立超热活化分析和循环活化分析方法 ,制备了医用放射性玻璃微球。在改善微堆运行性能的基础上 ,拓宽微堆的应用 ,摸索一条新的发展道路。     

医院中子照射器反应堆实验研究

医院中子照射器是专用于硼中子俘获治疗的核装置,所用反应堆功率为30 kW,采用~(235)U富集度为12.5%的UO_2为燃料,金属铍反射层,轻水为慢化剂和冷却剂.堆芯产生的热量靠自然循环冷却.在反应堆堆芯相对两侧分别设置了热中子束流和超热中子束流,用于治疗患者.在微堆零功率实验装置上,完成了临界质量、控制棒效率、上铍反射层效率及其它部件反应性的测量,确定了最终燃料元件的装载,为工程物理启动提供实验数据.     

我国大型高通量工程试验堆建成

我国自己设计制造的大型高通量工程试验核反应堆,已经在二机部第一研究设计院建成。反应堆已于1980年12月按第一炉装载的预定参数投入高功率运行。这座反应堆是压力壳型,铍水慢化,铍作反射层的高通量工程试验反应堆。反应堆设计热功率为12.5万千瓦。活性区内有φ150的辐照孔道5个,φ63的辐照孔道2个。反射层内有φ230和φ120的辐照孔道各2个,材料辐照罐和同位素辐照靶件可在栅格上任意布置。     

微型中子源反应堆改进方案研究

分析研究了使微型中子源反应堆延长单次运行时间的几种改进方案。研究与实验结果表明,在中心控制棒与堆芯原有结构保持不变,不超过安全运行限值的情况下,在侧铍反射层外附加弧形补偿板,再添加适当当量的顶铍片后,中心控制棒与补偿板联动,可使微型中子源反应堆一次连续运行时间大大提高．提高了微型中子源反应堆的运行性能。     

开发微堆潜力的设想

微堆存在二个潜在资源 ,即附加补偿棒和池水。在确保安全的前提下 ,可将它们挖掘并加以利用。设置附加补偿棒 ,把闲置的顶铍反射层的后备反应性加以利用并逐步释放。微堆为“罐 池”结构 ,运行时堆水与池水存在较大温差 ,2 7t池水可作为一大的冷源 ,通过池水中的螺旋冷却器冷却 1 4t堆水 ,可获得较多的温度效应。采取了这两项措施之后 ,可大大延长微堆的可连续运行时间 ,改善它的运行性能 ,拓宽了微堆的应用 ,提高其寿期内的价值     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

核电厂压水堆围板对堆芯中子经济的影响

本文研究了围板对堆芯中子经济的影响,并且经过不同厚度围板的零功率反应堆的实验,验证了理论予计的结果。尽管不锈钢通常被认为是一种寄生性中子吸收体,但发现围板增加到某一厚度时,反而提高了堆芯的中子经济效益。由于不同厚度不锈钢板反射中子的作用不同,当围板厚度由2.5cm(这是核电厂压水堆使用的一种代表性围板的厚度)增加到20cm(在结构设计中应将20cm的围板改变为不锈钢反射层组件)时,则30万kW的核电厂压水堆第一燃料循环的寿期将会延长40EF-PD。若把第一循环寿期转换为平衡燃料循环寿期(360EFPD),20cm不锈钢反射层则可多发33EFPD。再按40年的核电厂设计寿命(年负荷因子为0.8)折算,则可多发76亿度的电。如果不锈钢反射层用锆合金代替,其经济效益更大。     

零功率微堆的蒙特卡罗数值模拟计算

本文用蒙特卡罗程序MCNP/4B计算了中国原子能科学研究院零功率微堆的堆芯物理参数、铍反射层价值和中子实验孔道的部件反应性价值等数值,并与实验测量值比对.结果表明,两者符合一致,验证了理论计算模型的可靠性.     

高通量工程试验反应堆

我国高通量工程试验反应堆(HFETR)是一座压壳型反应堆,它采用高浓铀套管元件,水作慢化剂和冷却剂,铍作反射层,热功率125兆瓦,燃料内最太热中子通量6.2×10~(14)中子/厘米~2·秒。该堆已于1980年12月16日高功率运行。     

商用微堆安全特性的研究

文章描述了商用微堆反应性温度系数在零功率实验装置、商用微堆稳态运行时和引入不同反应性的暂态试验中的相应结果。文中给出了有关试验的结果图表,将这些图表的数据与原型微堆的有关数据进行比较,可以得出商用微堆的安全特性优于原型微堆的结论。     

微型反应堆照射座内热中子通量谱的测定

一、基本原理用一组展开函数?_i(E)来表示所测的真实谱φ(E),典型的展开函数是一组N—1项的多项式,N是探测箔种类数。     

微型反应堆辐照座内中子温度和超热指标的测定

一、引言对于高浓铀燃料、金属铍反射层,主要作为中子活化分析用的微型反应堆而言,对有关辐照座内的能谱和谱参数必须有所了解,中子温度是重要的谱参数,它基本上反映了反应堆热谱的特征。     

Xenon poisoning calculation code for miniature neutron source reactor（MNSR）

In line with the actual requirents and based upon the specific characteristics of MNSR,a revised point-reactor model was adopted to model MNSR‘s xenon poisoning.The corresponding calculation code.MNSRXPCC(Xenon Poisoning Calculation Code for MNSR),was developed and tested by the Shanghai MNSR data.     

微型反应堆的核安全审评与监督

前言功率为27kW、堆芯大小为直径×高度=242×250mm~2,以含富集度为90.3%~(235)U的铀铝合金为燃料元件,中子通量密度达1×10~(12)n/cm~2·s的微型反应堆,主要用于中子活化分析、短寿命同位素生产、教学和培训等。     

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.     

MNSR Annual Operation Report

Guided by the nuclear Safety roles, code and Standards of China, MNSR performed safe operation in 2006. The annual operation report of MNSR in 2006 is as follows.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

Thermal hydraulic analysis of Syrian MNSR research reactor using RELAP5/Mod3.2 code

This paper presents the development and validation of a MNSR-RELAP5 model. MNSR is a 30 kW, light-water moderated and cooled, beryllium-reflected, tank in pool type research reactor. A RELAP5 model was set up to simulate the entire MNSR system. The model represents all reactor components of primary and secondary loops with the corresponding neutronic and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents are simulated.