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。     

秦山核电厂压水堆考验燃料组件辐照后检验

为验证秦山核电厂压水型反应堆燃料组件的设计、制造工艺和材料性能，采用３×３考验组件，在原子能院重水试验堆ＨＷＲＲ的高温高压回路中模拟秦山核电厂反应堆的稳态和短时超功率工况进行了综合考验，平均燃耗达２５７００ＭＷｄ／ｔＵ。对考验组件和燃料棒作了综合性的辐照后检验，检验项目包括：燃料棒的外观观察、尺寸测量、γ扫描、涡流探伤、Ｘ射线照相、裂变气体释放率测量，包壳管、控制棒导向管和格架弹簧片的力学性能试验，包壳和燃料芯块的微观组织分析和定量测量，水垢的Ｘ射线衍射分析等。检验结果表明：考验组件设计合理，制造工艺可靠，燃料芯块、包壳和其它材料的性能均能满足要求。所取得的检验结果可为秦山核电厂压水堆的燃料组件以及同类燃料组件的设计、制造和性能改进提供依据。     

CARR堆标准燃料组件堆外水力稳定性试验

CARR采用平板型燃料组件,燃料板采用6061铝合金包壳、U3Si2-Al弥散型燃料,燃料板与侧板用滚压法固定,形成一个整体。除辐照条件外,在模拟堆内运行热工水力及水质条件下,对标准燃料组件减小水力稳定性试验,试验流速为设计流速的120％。第1个组件试验中发现的问题,及时反馈给设计方和制造厂家,对经过设计和制造工艺改进后制造的第2个组件,又进行了试验。两     

新型U3Si2-Al燃料板的正电子湮没寿命谱研究

用正电子湮没寿命谱仪对U3Si2-Al燃料板样品的正电子湮没寿命进行了测量及分析,得到不同工艺状况下燃料包壳材料微观缺陷的形态及变化趋势.回火态燃料板包壳基体中的微观缺陷以单空位的点缺陷为主;冷作态中的缺陷以双空位、位错等缺陷为主;冲刷态中的缺陷以层错、小的空位团等缺陷为主.3种样品中,均未发现影响燃料板安全的大空位团缺陷.回火和冲刷等工艺或运行工况,会使燃料板包壳基体中的微观缺陷发生转变,并改变了燃料板的宏观力学性能.     

燃料元件辐照后破坏性检验

介绍了现在通行的燃料棒辐照后破坏性检验的一般方法,包括燃料辐照后的微观结构观察和力学性能试验;描述了光学显微镜,透射电子显微镜和扫描电子显微镜在观察燃料微观组织结构时的应用;论述了辐照后包壳管在热室中的拉伸、爆破、蠕变等试验方法,比较辐照前后包壳力学性能的变化.这些方法有利于加深理解燃料芯块、包壳和结构部件的辐照行为, 为提高燃耗提供依据.     

乏燃料废包壳残留物质核素分析计算

为分析计算乏燃料废包壳残留物质的核素含量，以M310型核电机组及燃料组件为分析对象，建立了乏燃料废包壳残留物质核素含量分层计算模型，用SCALE程序计算分析了²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比等主要参数随燃耗及冷却时间的变化。计算结果表明，²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比随燃耗及冷却时间的变化均可用三阶多项式拟合。本文工作为废包壳残留物质非破坏性测量方法研究提供了数据支持。     

热蠕变对UMo/Zr单片式燃料板起泡行为的影响

针对含有气腔的UMo/Zr单片式燃料板,考虑包壳材料的热蠕变效应,将包壳的变形与气腔压力相耦合,发展了一种对燃料板宏观起泡行为进行数值模拟的方法。基于所建立的模拟方法,计算分析了包壳热蠕变和气腔内裂变气体原子数对起泡行为的影响。研究发现,在考虑包壳热蠕变时,若局部开裂区域内的裂变气体原子数为4.0×10¹⁷,以鼓泡高度0.1 mm作为起泡阈值的判断标准,所预测出的阈值温度比不考虑热蠕变时低100℃;若局部开裂区内的裂变气体原子数由2.5×10¹⁷增加至4.0×10¹⁷,则燃料板的起泡阈值温度将可能降低40℃,通过降低包壳材料的热蠕变率可以有效提高燃料板的抗鼓泡能力。     

研究堆用U3Si2-Al燃料板的热稳定性试验

U3Si2-Al板状燃料组件是一种推广应用的新型燃料元件,在国内首次应用。燃料组件的各项性能,特别是热稳定性必须通过实验验证。通过对铀密度为3.02 g/cm3的U3Si2-Al燃料板的热稳定性试验,得到:热稳定性试验会使燃料板的体积略有增大;120℃及250℃的热循环下,燃料板无明显变形,表面无变化,400℃的热循环下,燃料板略有弯曲,个别芯体裸露的燃料板表面有起泡现象;循环温度越高,芯体中U3Si2颗粒开裂越严重等实验结论,为该燃料组件的结构设计、安全分析、加工工艺提供了关键参数,并为该组件的堆内运行提供了借鉴。     

基于流固热耦合的燃料元件包壳结构完整性分析

反应堆系统发生瞬态工况时,冷却剂温度的瞬间大幅度变化会对燃料元件包壳结构完整性造成冲击,危及反应堆安全。本文以某压水堆3×3燃料组件为对象,采用流固热耦合方法对冷水事故下燃料组件的流动换热特性和燃料元件包壳温度、变形及应力进行了三维精细化模拟。结果表明:定位格架能够增强燃料棒表面的对流换热强度;包壳变形时向与刚凸接触的一侧折弯,向与弹簧接触的一侧凸起;包壳与定位格架接触部位的温度和最大等效应力随事故时间不断增大,且最大等效应力超过了包壳材料的屈服强度,将发生强度失效,影响其结构完整性。本文研究可为反应堆燃料元件包壳瞬态工况下的完整性评价提供借鉴。      

俄继续开展耐事故燃料测试

正【英国《国际核工程》网站2021年5月4日报道】俄罗斯博奇瓦尔无机材料研究所(VNIINM)2021年4月29日宣布计划在年底完成耐事故燃料试验组件的第三个辐照周期测试。俄2019年1月在核反应堆研究所(RIAR)MIR研究堆中启动对首批两个耐事故燃料试验组件的辐照测试。两个组件由新西伯利亚化学浓缩厂(NCCP)制造,含有2种燃料芯块和2种包壳:燃料芯块分别是传统二氧化铀芯块和具有更高铀密度和导热性的铀钼合金芯块;包壳分别是带铬涂层的锆合金包壳和铬镍合金包壳。     

A parametric study of multichannel molten cladding motion under unprotected loss-of-flow accident conditions in LMFBR's

One of the key assumptions of the present multichannel clad motion model was that the total pressure drop over the voided channel could be supplied as a boundary condition. The incoherency effect on cladding motion can be significant for a full-scale subassembly, and therefore parametric studies of the total pressure drop and oscillatory pressure effect due to sodium chugging were examined using the multichannel model.There is an axial blanket region in demonstration plant or commercial-power-plant designs instead of a reflector in FFTF design above the top of fuel. It was shown that due to the difference in the thermal conductivities between the blanket material and reflector, significant changes in the timings of various events of the cladding relocation might occur. It is also noted that depending on the effect of the sodium voiding on the reactivity, the fuel may become molten when the molten cladding is still around. The possibility of the occurrence of this situation is studied by increasing the power in the present model.     

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。     

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.     

铝基碳化硼堆内辐照热工安全分析

铝基碳化硼是一种新型的乏燃料贮存架结构材料,需对其各项性能进行研究,其中,铝基碳化硼材料的耐辐照性能是关键参数之一。为进行铝基碳化硼材料的堆内辐照考验,并保证其在堆内辐照的安全,针对铝基碳化硼辐照方案的特点,采用了CFD程序进行热工校核计算,分析了铝基碳化硼材料在堆内辐照的安全特性,优化了堆内辐照方案。     

Modeling thermal conductivity in UO2 with BeO additions as a function of microstructure

New processing methods show promise for improved thermal conductivity in UO₂ by the incorporation of a highly-conducting material. Such composites are likely to have anisotropic microstructures which bring new challenges to thermal conductivity simulation but also significant potential for improvement in the thermal performance. This paper presents simulation results for the thermal conductivity of UO₂/BeO composites using statistical continuum mechanics. The results successfully capture the microstructural heterogeneity and predict the corresponding anisotropic thermal properties. The application of statistical continuum mechanics to materials design makes it possible to design novel anisotropic fuel pellets with enhanced thermal conductivity in a preferred direction.     

Design and validation of breeder unit of helium-cooled pebble bed blanket for demo fusion reactor

In the framework of European helium-cooled pebble bed (HCPB) blanket development, an HCPB breeder unit based on the design of pebble beds between flat cooling plates is proposed for a DEMO fusion reactor. The performances of the designed breeder units are validated by supporting analyses. By applying the thermal boundary conditions obtained by neutronics simulations for the DEMO reactor, results of finite element calculations of the breeder unit are analyzed in views of thermal-hydraulics and thermal stress to identify the adherence to maximum temperatures in structural and functional materials and the abidance by the stress criterion imposed by the structural material. The layout of the internal meandering channels in the cooling plates is optimized by using numerical methods. Finally, possible improvements of the new designed breeder unit are proposed.     

聚变增殖堆液态金属Li自冷包层的热工水力分析

完成了托卡马克商用混合堆 TCB(Tokamak Commercial Breeder)Li 自冷包层设计的热工水力分析,讨论了热工水力设计中的一些关键问题。用两维有限元热传导程序 AYER 计算了 TCB 包层的温度分布,用液态金属 MHD(Magnetohydraudynamic)压降公式计算了包层的压降。同时,还分析了包层冷却剂丧失事故 LOCA 的瞬态热工过程。分析表明,正常工况下,包层结构材料最高温度,结构材料与冷却剂界面最高温度,以及包层总压降都满足堆设计要求。在 LOCA 工况下,如果停堆后1小时内包层中的燃料球能够借助重力卸出包层,第一壁和包层是安全的,并且不会受到损伤。     

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.