微型反应堆设计、运行经验总结

微型反应堆从原型微堆到商用微堆走过了十多年的历程。在设计、运行方面积累了丰富的经验，集中到一点就是如何处理好经济性与安全性这一对矛盾，即既要使建在人口稠密地区的微堆，确保其安全，不会发生任何的核事故，又要在有限大小的铍环反射层内，选择合适的氢原子和铀２３５原子比例的栅元，使设置在铍环反射层中的辐照孔道内，由较低的堆功率获得较高的中子通量密度，尽可能获得长的运行时间和炉寿期。一般反应堆炉寿期较短，经过１－２年就换料。然而微堆的炉寿期有２０－３０年。制约微堆炉寿期的主要因素不是后备反应性，而是核燃料元件包壳的腐蚀速率，如何监测微堆微量的核泄漏、防止其周围环境不被污染是微堆运行过程中一个突出问题     

微型核反应堆及其在中子活化分析中的应用

上海测试技术研究所的微型反应堆,是中国原子能科学研究院设计制造的国内第3座商用微堆,于1991年12月18日达到临界,1992年1月7日投入为期一年的试运行;至1992年12月12日共开堆101天,累积的开堆功率为8074,2 kW·h,每根内辐照座处的积分中子通量     

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。     

尼日利亚微堆低浓化的设计计算和实验验证

为了顺利开展低浓铀转化，使用MCNP、WIMSD和CITATION程序建立了尼日利亚微堆(低浓铀)模型，计算了尼日利亚微堆(低浓铀)的几个主要参数，并在微堆零功率装置上完成了零功率实验，测量获得中心控制棒、上铍、最外围元件、内辐照管和外辐照管等实验部件的反应性效率，实验数据和计算结果符合较好。完成了尼日利亚微堆(低浓铀)现场安全功能实验，实验结果证实尼日利亚微堆(低浓铀)具有良好的安全性能，标志着中国原子能科学研究院完成了尼日利亚微堆(低浓铀)低浓铀转化。     

巴基斯坦微堆内辐照座通量密度与堆功率的测定

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。     

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。     

巴基斯坦微型堆启动与参数测定

文章叙述了本院为巴基斯坦核科学技术研究所建造的微型堆(PARR-Ⅱ)的启动、调试与参数测量工作的主要结果。该堆于1989年11月2日达到临界,初始冷态的后备反应性为4mk,一周后提升功率,内辐照座中子通量密度达到额定值1×10~(12)n/cm~2·s,在1.1×10~(12)n/cm~2·s通量密度水平最大可连续运行时间为6.75h。通过释放4mk反应性实验,在内辐照座测量得到中子通量密度达最大值时为额定通量密度的2.9倍。测试结果表明,该堆具有良好的固有安全性,各项技术指标均达到了预期值,并于1989年11月下旬正式投入营运。     

微型反应堆燃料低浓化可行性初步研究

应用蒙特卡罗计算程序,模拟计算山东微堆的堆芯参数,包括keff、βeff、控制棒价值、上铍效率、内辐照中子通量以及停堆深度,计算结果与实验结果基本一致.保持微堆堆芯尺寸不变,采用低富集度UO2芯体燃料棒替换原来的高浓铀燃料棒,计算不同235U富集度下微堆的有效倍增系数keff,据此确定满足要求的UO2富集度为12.5%.在此基础上计算了富集度为12.5%的低浓堆芯参数,并与高浓堆芯参数进行了比较.结果表明,微堆燃料低浓化是可行的.     

核径迹热释中子探测器测量高中子通量密度

本文提出一种用于高中子通量密度测量的方法,即使用核径迹热释中子探测器测量中子通量密度,该方法在低中子通量密度测量方面已成功在微型中子源反应堆上得到验证。为了测试其在高中子通量密度测量方面的适用性,在中国先进研究堆辐照孔道内进行了应用研究。结果表明:孔道内中子通量密度相对分布总体趋势与MCNP的计算结果符合较好,此种方法测量高中子通量密度有效可行。     

原型微堆八年运行及进展

1 综述由中国原子能科学研究院研制的微型中子源反应堆的原型堆(简称原型微堆)于1984年3月9日首次达到临界,3月31日达到额定功率。迄今,原型微堆已安全运行8年多,为商用微堆的改进设计和建造积累了丰富的实验验证数据。同时,微堆还做了大量的中子活化分析工作(涉     

Feasibility study of photo-neutron flux in various irradiation channels of Ghana MNSR using a Monte Carlo code

Computer simulation was carried out for photo-neutron source variation in outer irradiation channel, inner irradiation channels and the fission channel of a tank-in-pool reactor, a Miniature Neutron Source Reactor (MNSR) in sub-critical condition. Evaluation of the photo-neutron was done after the reactor has been in sub-critical condition for three month period using Monte Carlo Neutron Particle (MCNP) code. Neutron flux monitoring from the Micro Computer Control Loop System (MCCLS) was also investigated at sub-critical condition. The recorded neutron fluxes from the MCCLS after investigations were used to calculate the power of the reactor at sub-critical state. The computed power at sub-critical state was used to normalize the un-normalized results from the MCNP.     

Four energy group neutron flux distributions in the Syrian Miniature Neutron Source Reactor using the WIMSD4 and CITATION codes

A 3-D (R, θ, Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the pointwise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation sites with relative differences less than 7% and 5%, respectively.     

On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.     

Design of a permanent epithermal neutron irradiation site in the Syrian MNSR

A permanent epithermal neutron irradiation site was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using cadmium as a thermal neutron shielding material. This site was designed by Cd-shielding the internal surface of the outer aluminum tube of the FOIS (First Outer Irradiation Site) in the MNSR. The MCNP-4C calculations showed that, to have a permanent epithermal neutron irradiation site for the ENAA using the cadmium, it is necessary to add the top beryllium shims of the reactor to compensate for the reactivity losses due to the neutrons absorption in the cylindrical cadmium shell. The activation detectors were used to measure the thermal and epithermal neutron fluxes in the FOIS. Distribution of the thermal neutron flux along the vertical direction of the outer irradiation capsule used in the FOIS has been found using MCNP-4C code, and experimentally by irradiating five copper wires. Good agreements were obtained between the calculated and the measured results.     

医院中子照射器中子束流出口处热中子注量率的测量

医院中子照射器是基于微型反应堆而设计的专门用于硼中子俘获治疗（BNCT）的核反应堆装置,其额定功率为30 kW。在堆芯相对两侧分别设有一条热中子束流和超热中子束流用于病人照射,在热中子束流内引出一条实验用热中子束流,用于瞬发γ法测量病人血硼浓度。本工作利用²³⁵U裂变靶和白云母探测片测量了热、超热和实验用热中子束流出口处的热中子绝对注量率。结果显示,在30 kW额定功率运行时,热、超热和实验用热中子束流出口处的热中子注量率分别为1.67×10⁹、2.44×10⁷和3.03×10⁶ cm^-2•s^-1。以上结果达到了BNCT设计要求,并能满足瞬发γ测量血硼浓度的要求。     

Comparative assessment of theoretical and experimental measurement of photo neutron flux level in GHARR-1 after shutdown

Direct photo-neutron source strength was evaluated for the Miniature Neutron Source Reactor (MNSR) in subcritical condition in the GHARR-1 facility. Two different static methods were applied for comparison. A theoretical method based on the use of MCNP code and an experimental method based on foil activation technique. The latter has been found to be most convenient method for neutron flux measurement. The method depends only on the activity of a bare and cadmium covered foil if the irradiation positions are known. Photo-neutron flux level was determined theoretically using MNCP after measuring neutron flux at shutdown; and experimentally using Neutron Activation Analysis (NAA) technique also at shutdown with great care. The values obtained from the theoretical and experimental measurements are tabulated in Table 2. The results recorded were validated using biological peach leave and a geological rock sample. The results after validation for Mn concentration in the samples were 87 ± 1 μg/g and 432 ± 23 μg/g, respectively. Results for the two methods were in good agreement. Realization of photo-neutron source existence due to beryllium reflector was also experienced.     

A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP

A neutronics feasibility study has been performed to determine the enrichment that would be required to convert a commercial Miniature Neutron Source Reactor (MNSR) from HEU (90.2%) to LEU (<20%) fuel. Two LEU cores with uranium oxide fuel pins of different dimensions were studied. The one has the same dimensions as the current HEU fuel while the other has the dimensions as the special MNSR, the In-Hospital Neutron Irradiator (INHI), which is a variant of the MNSR. The LEU cores that were studied are of identical core configuration as the current HEU core, except for potential changes in the design of the fuel pins. The following reactor core physics parameters were computed for the two LEU fuel options; clean cold core excess reactivity (ρ_ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, β_eff and prompt neutron lifetime, l_f). Results obtained are compared with current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of NIRR-1 in particular from HEU to LEU.     

嫦娥五号月壤中子活化分析研究

我国嫦娥五号从月球风暴洋东北部、人类此前从未去过的月球中纬度地区(51.916°W,43.058°N)采集回了新的月壤样品,弥补了40多年前美国和苏联采样区域有限的不足,拓展了月壤样品的代表性,具有重要的科研价值.中子活化分析具有高灵敏、多元素、非破坏等优点,基于中国原子能科学研究院49-2泳池堆和微型中子源反应堆,利...