中国先进研究堆标准燃料组件堆外水力稳定性试验

中国先进研究堆（CARR）标准燃料组件由滚压在两块侧板上的21块燃料板组成。堆外水力试验的目的是考验在水力冲刷条件下燃料组件的结构稳定性。试验件是按照正式产品制造工艺制造的贫铀组件，试验平均流速为12m/s，是满功率运行流速的120％。先后试验了2个组件，第1个组件试验60d，是满功率运行时间的120％，试验后观察到固定下定位梳的销钉松动，下定位梳严重磨损了燃料板;工艺改进后制造的第2个组件试验120d，是满功率运行时间的240％，试验表明，第2个组件结构完整。试验中对组件结构稳定性和燃料板腐蚀性能，诸如组件的压差、燃料板振动、包壳表面腐蚀深度等进行了研究。     

板型燃料组件额定流速流致振动实验研究

针对贫铀叠层板型燃料组件,采用了一种新的软测量方法,进行了其在经受冷却剂额定流速冲刷时的振动实验。获得了该组件在额定流速6 m/s下的流致振动动态响应时间域和频率域结果,并对实验结果进行了分析。     

板状燃料组件非线性振动特性

设计1个能模拟堆芯燃料组件工作环境和力学边界条件的台架,研究板状燃料组件的非线性振动行为。对安装在振动台上的系统进行白噪声随机激励,得到燃料组件响应-激励关系不同方向、不同介质、不同振级共42种工况的系列曲线。试验结果表明:随着激励振级加大,板状燃料组件的第一阶共振频率减小;板状燃料组件在水中的第一阶共振频率远比在空气中的小,且非线性特性仍呈软弹簧特性;板状燃料组件在环境中的振动阻尼大,能大大抑制抗震响应。     

破损燃料组件热室检查技术研究

燃料组件破损直接影响了反应堆的安全运行，分析燃料组件破损原因是燃料组件研发改进的重要环节。通过破损燃料组件水下解体、破口位置定位、破口试样取样等关键技术的研究，建立了破损燃料组件热室检查方法。研究结果表明，该技术路线合理，检查方法可行，为热室条件下开展燃料元件破损检查提供了技术途径。?      

板状燃料组件流量分配CFD研究与优化

板状燃料组件被广泛应用于研究堆中,组件内的流量分配是设计时需要考虑的一项重要内容。计算流体动力学(Computational Fluid Dynamic,CFD)方法是研究流量分配的重要手段,但有限的计算资源限制了其在板状燃料组件流量分配研究中的推广。针对板状燃料组件冷却剂流道狭长、封闭的特点,提出了部分建模迭代求解的计算方式,将无流量分配组件与有流量分配组件两种工况下各流道流量的计算值与直接完整建模的结果进行了对比,最大误差分别为0.56%与0.81%。鉴于前者对计算资源的需求远小于后者,部分建模迭代求解可以作为板状燃料组件流量分配CFD研究的合理可信的优化方案。     

堆芯燃料组件流阻特性模块式与整体式三维数值模拟方案比较

组件的阻力特性影响堆芯不同类型组件的流量分配，对堆芯的设计起到至关重要的影响。为提高验证堆芯燃料组件特性的求解精度及效率，本文针对燃料区6类燃料组件中的两类进行模块式及整体式三维数值模拟，获得了两类组件的流阻特性，并用相同条件下的全组件试验结果进行了对比。结果表明：推广至堆芯所有燃料组件流阻特性预测，模块式所需计算时间约为整体式的1/6，但整体式三维数值模拟所得压降与试验结果吻合度高，误差较模块式小。最后深入研究了流速及温度变化对流阻特性的影响。该研究为后续各类组件的流阻特性研究方法选取提供技术支持。     

中国先进研究堆(CARR)工程燃料组件设计验证试验

中国先进研究堆(CARR)采用的燃料组件在国内尚属首次加工与使用。为了保证燃料组件的完整性和安全性,满足堆安全运行的需要,对燃料板和组件的结构稳定性、流致振动、临界流速、热循环、堆内辐照等进行了设计验证试验。结果表明,CARR燃料组件的设计和加工工艺是合理的,谈组件在反应堆实际运行条件下是稳定和安全的。     

新型U3Si2-Al燃料组件的工程验证试验研究

新研制的U₃Si₂-Al板状弥散型燃料组件结构复杂,国内对该燃料组件的结构材料、制造工艺、力学性能、运行特性等均缺少经验及评定标准。为得到该新型燃料组件的各种性能参数,开展了燃料包壳及结构材料的力学性能试验、燃料板及包壳材料的热物性及热稳定性试验、燃料板的力学性能试验、燃料板的正电子湮灭寿命试验、燃料组件的水力冲刷和解体试验等一系列的工程验证试验和专项研究,得到的各项实验数据为燃料组件的结构设计、可靠性分析、安全审评提供了重要依据,也为燃料组件的加工制造、堆内使用管理提供了借鉴。     

板状燃料组件窄缝流道流速分布实验研究

实验研究了平行窄缝流道板状燃料组件的流速分布。实验以无离子水作流动介质，在常压及不同Re条件下进行。实验结果表明：在不同Re条件下，横向单通道流速分布呈中间流速高、边缘流速陡减的梯形分布，符合常规流道流速分布特点；纵向平行多通道间的流型呈中间流速低、边缘流速高的凹型分布，这种凹型分布是由入口结构件造成的，对反应堆堆芯流量的均匀分配不利。     

反应堆燃料组件解体技术研究

充分考虑反应堆燃料组件结构特点,提出了一种先翻转、再拆卸下管座、最后拔取燃料棒的解体工艺,并设计了与解体工艺相对应的专用工具。采用集成化的思路,使所有设备均布局在投影面积仅为1.2 m~2的面积以内,既满足了现场安装条件的限制,又保护了乏燃料水池已有设备。采用该技术顺利完成了水下6 m处反应堆燃料组件的解体工作。     

CEFR堆芯燃料组件抗震试验方法研究

为了验证中国实验快堆（CEFR）堆芯燃料组件的抗震性能,保证地震下结构完整性和气密性,必须研究制定兼具代表性和包络性的堆芯组件抗震试验方法。本文基于俄罗斯组件耐振试验方案分析,结合国内试验规范和堆芯实际约束条件,提出了一套新的组件抗震试验方法,并通过分析计算论证新方法的合理性。结果表明：新方法的试验结果是保守的,可保证在相同地震输入下单组件应力、冲击响应基本能包络处于堆芯组件群中的组件响应,新方法要求单根组件分别在刚性台架和柔性台架上依次完成抗震试验。本文结果对快堆堆芯组件的抗震试验研究具有重要指导意义。     

中国实验快堆某辐照实验组件燃耗分布测量

在中国实验快堆（CEFR）上建立了实验组件燃耗分布测量的实验装置。对CEFR某一辐照实验组件中的4#及6#燃料元件棒进行了相对燃耗分布的测量，并与理论计算结果进行了比较。结果表明：两根燃料元件棒虽处于实验组件的不同位置，但相对燃耗分布基本一致；燃耗分布的实验测量结果与理论计算结果符合较好；实验组件燃耗分布测量的相对误差在10.2%以内。本文工作为开展快堆乏燃料组件燃耗测量奠定了基础。     

Experimental qualification of subassembly design for Prototype Fast Breeder Reactor

The Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam, India, is a 500 MWe sodium cooled pool type reactor. The core of the PFBR consists of 1758 free standing subassemblies supported on the grid plate. The entire core is divided into 15 different flow zones and the flow rate required through each zone is calculated based on the fission heat generation. The coolant sodium flows from the bottom of the subassembly to top and the design of the subassembly for each flow zone is quite complex. There are 181 fuel subassemblies in PFBR core with 217 fuel pins in each subassembly, vertically held in the form of bundle within a hexagonal wrapper tube. The pins are separated by spacer wires wound around the pins helically. Analytical prediction of subassembly pressure drop, vibration and determination of inception of cavitation for this complex geometry is very difficult. So experiments were conducted extensively to get a more accurate evaluation of the design and for its qualification for the use in PFBR, which is designed for 40 years of operation.Pressure drop and cavitation experiments were carried out in water on full scale (1:1) subassemblies of all flow zones. The overall pressure drop of the subassembly determines the ratings of the pump. Cavitation of the pressure drop devices lead to erosion damage of fuelpins and may also result in reactivity fluctuation due to sodium-void effect. So it is essential to confirm that the subassembly is not cavitating in the operating regime of the reactor. Subassembly can vibrate in cantilever mode due to the turbulence in the flow and can result in reactivity fluctuation, reactor control problem and can even lead to the failure of the fuel pins. So vibration measurements were carried out in water on the maximum rated subassembly. This paper discusses various experiments carried out on PFBR subassembly, the similarity criteria followed, instrumentation, results and conclusion.     

Computational fluid dynamic investigations of partial blockage detection by core-temperature monitoring system of a sodium cooled fast reactor

Thermal hydraulic studies have been carried out to understand temperature dilution suffered by core-temperature monitoring system of a sodium cooled fast reactor. The three-dimensional computational model is validated against experimental results of a water model. Jet mixing phenomenon as predicted by different turbulence models is compared and RNG k–? model is found to be better than other models. A comprehensive parametric study considering: (i) effects of construction/manufacturing tolerances on thermocouple positions with respect to subassembly positions, (ii) thermal/irradiation bowing of subassemblies, and (iii) changes in core power profile during reactor operation cycles has been carried out. The studies indicate the maximum possible dilution in fuel and blanket subassemblies to be 2.63 K and 46.84 K, respectively. Shifting of thermocouple positions radially outward by 20 mm with respect to subassembly centers leads to an overall improvement in accuracy of thermocouple readings. It is also seen that subassembly blockage that leads to 7% flow reduction in fuel subassembly and 12% flow reduction in blanket subassembly can be detected effectively by the core-temperature monitoring system.     

Pressure pulse on a subassembly wall due to gas release from a failed fuel pin

This paper presents a mathematical model to predict the pressure pulse on a subassembly of fuel pins due to rapid release of gas from a failed pin into liquid coolant between the pins. The subassembly is simulated by a rigid circular tube, and liquid flow inside the tube is assumed incompressible, inviscid, and irrotational. A gas bubble along the centerline of the subassembly is considered to be formed as a result of the gas release from the plenum, and a pressure pulse on the subassembly wall is a consequence of the liquid being accelerated by the gas bubble. It is assumed that the gas bubble grows spherically until it touches the subassembly wall, and then expands as a cylinder with hemispherical ends. This analysis is particularly applicable to the EBR-II reactor.     

Probabilistic distributions of pin gaps within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to pin gap

Probabilistic fuel pin gap distributions within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to fuel pin gaps are discussed. The analyses consist mainly of expressing a local fuel pin gap in terms of sensitivity functions of the related uncertainties and calculating the corresponding probabilistic distribution through taking all the possible combinations of the distribution of uncertainties. The results of illustrative calculations show that with the reliability level of 0.9987, the maximum deviation of the pin gap at the cladding hot spot of a center fuel subassembly is 8.05% from its nominal value and the corresponding probabilistic pin gap distribution is shifted to the narrower side due to the external confinement of a pin bundle with a wrapper tube.     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

快堆燃料组件热工流体力学计算研究

对于钠冷快堆,在燃料和包壳最高温度等设计限值下,为获得较高的堆芯出口温度,需深入分析燃料组件内的热工流体力学问题,准确预测组件内的冷却剂温度分布。本文在CRT模型和F.C.Engel等人工作的基础上,提出了ICRT压降关系式,用以计算冷却剂在湍流区、过渡流区和层流区的棒束压降;引入CRT模型和WEST对流传热模型,改进了SUPERENERGY子通道分析程序,并将改进程序与原程序计算结果进行了对比,结果表明：最热子通道出口温度略有降低,液膜温压略有增加;并用计算流体力学软件CFX对中国实验快堆单盒燃料组件活性段进行了三维数值模拟,将计算结果用CRT模型、ICRT压降关系式及改进后的SUPERENERGY子通道分析程序进行了验证,相互符合较好。     

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.