摘要：据《Теплоэнергетика》2013年7月刊报道,核电站的建造最优先考虑发电的高效率（约50%）。在这方面,配备氦气轮机装置和氦气载热介质高温反应堆的核装置很有前途。机械制造设计局的专家对高温气冷反应堆核电站紧凑型热交换设备进行了成功的设计和详细的研究。将高温气冷反应堆和氦气轮机装置一起组成一回路的核电站设计必须以利用氦气循环高效的热交换设备为先决条件。提供了最佳热交换表面选择的分析和计算-试验研究的结果。并分析研究了回热器、端部冷却器和中间冷却器构造、形式及特点。

摘要：美国280MW的Mojave槽式光热电站于12月1日正式投运，Abengoa公司开发的Mojave槽式光热电站正式并网实现商业化运行，该项目位于洛杉矾东北方向100英里，装机280MW，净装机为250MW，年发电量可满足9万普通家庭的日常用电需求，年减排二氧化碳达35万吨。该项目采用常规的导热油槽式技术，且未配置储热系统。此前投运的Solana光热电站装机也为280MW，位于亚利桑那州，储热时长达6h。对这两个电站的实际运行效果进行对比可为储热经济性研究提供蓝本。

摘要：本刊讯 国家“863”计划的重大项目──由清华大学核能技术设计研究院负责设计和建造的我国首座高温气冷反应堆,日前建设进展顺利,设备的安装调试正在进行中,计划将在明年建成并投入使用。 高温气冷反应堆与目前我国秦山、大亚湾核电站的压水堆不同,它是一种固有安全性好的先进反应堆堆型,发电效率可达45％左右,在我国长远能源系统中具有广阔的应用前景。这种反应堆是唯一能提供高温核工业热的核能源,它能产出950℃以上的高温,可广泛应用于发电、供热、稠油热采、煤炭气化液化等。由于高温气冷反应堆具有在事故情况下能自动停堆,剩余发热可以非能动地排除到堆外等一系列安全保障特点,它可建在工业区内或人口稠密的城市附近,以便更有效地服务于社会生产和人民生活。

摘要：介绍了１０ＭＷ高温气冷反应堆（ＨＴＲ－１０）位于反射层中的控制棒反应性当量的计算方法。用ＧＡＭ和ＴＨＥＲＭＯＳ程序分别产生堆芯、反射层、含硼碳砖及控制棒组成材料的超热群和热群截面。用二维离散纵标法程序ＳＮ２Ｄ在（ｒ，θ）坐标系下作详细控制棒结构的模型计算，该模型包括堆芯、反射层及反射层外的含硼碳砖，控制棒位于反射层中。含硼碳砖的外表面为自由边界，以考虑反射层中的中子泄漏谱。按通量权重归并控制棒区（包括控制棒、空隙及石墨反射层的整个圆环）的均匀化截面。全堆有控制棒和无控制棒情况下的Ｋｅｆｆ本征值，是由有限差分程序ＣＩＴＡＴＩＯＮ在（ｒ，ｚ）坐标系下计算出的，并由此得到控制棒的反应性当量。文中给出了ＨＴＲ－１０各运行阶段（包括初装堆、过渡过程中期和后期、平衡换料等时期）的控制棒的反应性当量。初装堆控制棒的反应性积分与微分当量也在文中给出。

摘要：全胶凝工艺已经发展成为制造１０ＭＷ高温气冷堆燃料芯核，ＵＯ２微球的优选工艺；本文通过批量试验和工艺参数的控制，实现产品质量和产率的控制。批量试验结果表明：藉助于分散频率和胶凝条件的控制，实现了微球尺寸和球形的控制：藉助于烧结参数，其中包括氢气氛、温度和时间等的控制，能够实现微球密度和Ｏ／Ｕ比的控制。在批量试验中微球产品的产率达到９０％。

摘要：为满足 10MW高温气冷堆 (HTR 10 )控制系统分析设计的需要 ,分析了HTR 10螺旋管式直流蒸汽发生器内两相流动和传热的实际过程 ,根据建模的要求和可能获得的结构数据 ,选择四方程漂移模型来描述螺旋管内两相流现象 ;经过仿真试算与实验总结 ,确定了模型中关键的结构关系式和传热工况的判断逻辑 ,以此为基础 ,建立了可反映直流蒸汽发生器在正常功率运行范围内动态特性的动态数学模型。通过对典型工况的动静态仿真计算和研究分析 ,表明所得的结果与理论定性分析和实验吻合 ,从而证明了该模型的正确性。

摘要：Small long-life transportable high temperature gas-cooled reactors(HTRs) are interesting because they can safely provide electricity or heat in remote areas or to industrial users in developed or developing *** paper presents the neutronic design of the U-Battery,which is a 5 MWth block-type HTR with a fuel lifetime of 5–10 *** a reactor pressure vessel diameter of less than 3.7 m,some possible reactor core configurations of the 5 MWth U-Battery have been investigated using the TRITON module in SCALE *** neutronic analysis shows that Layout 12×2B,a scattering core containing 2 layers of 12 fuel blocks each with 20% enriched235U,reaches a fuel lifetime of 10 effective full power years(EFPYs).When the diameter of the reactor pressure vessel is reduced to 1.8 m,a fuel lifetime of 4 EFPYs will be achieved for the 5 MWth U-Battery with a 25-cm thick graphite side *** 6×3 and 6×4 with a 25-cm thick BeO side reflector achieve a fuel lifetime of 7 and 10 EFPYs,*** comparison of the different core configurations shows that,keeping the number of fuel blocks in the reactor core constant,the annular and scattering core configurations have longer fuel lifetimes and lower fuel cost than the cylindrical ***,for the 5 MWth U-Battery,reducing the fuel inventory in the reactor core by decreasing the diameter of fuel kernels and packing fraction of TRISO particles is more effective to lower the fuel cost than decreasing the 235U enrichment.

摘要：Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor(HTGR).To solve the problem,a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the *** apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the *** designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field.A typical experimental process was analyzed in the paper,which lasted 76 hours including seven *** results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and *** temperature in the apparatus could be elevated up to 1600oC,which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor.