Hydraulics of Coolant Flow through Control Rod Channel in Very High Temperature Gas Cooled Reactor

Flow rate distribution and total pressure loss of a coolant flow through a control rod channel in the Very High Temperature Gas Cooled Reactor (VHTR) were analytically and experimentally examined. Helium gas of atmospheric temperature was used in the experiment; and the total mass flow rates ranged 0.005～0.05 kg/s and the gas pressures ranged 0.14～0.42 MPa. Pressure losses and flow rates in the control rod channel were measured. An analysis was made by using a one-dimensional flow network model for the inner and outer channels and the gap. The analytical results agreed fairly well with the experimental results on the flow rate distribution and the total pressure loss in the control rod channel.     

Startup Thermal Analysis of a High-Temperature Supercritical-Pressure Light Water Reactor

The startup systems of a high-temperature supercritical-pressure light-water-cooled thermal reactor (SCLWR-H), in which the core outlet temperature is 500°C and downward-flowing water rods are used as moderators, are studied by thermal-hydraulic analysis. The thermal analyses are carried out for various startup phases and detailed procedures for these phases are investigated. In constant pressure startup system, the reactor starts at supercritical pressure. A flash tank and pressure-reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at subcritical pressure. A steam-water separator and a drain tank are required. The separator is designed by referring to those of supercritical fossil-fired power plants (FPPs). The maximum cladding surface temperature is restricted not to exceed the rated value of 620°C. The minimum flow rate is 25% for constant pressure startup and 35% for sliding pressure startup. Both constant pressure and sliding pressure startup systems are found feasible from thermal analysis. Because of lower flow rate than SCFR, of which the core outlet temperature is about 430°C, the component weight required is reduced in SCLWR-H. The sliding pressure startup system should be used to reduce the component weight and to simplify the plant system.     

Calculational Methods of Control Rod Worth for High Conversion Light Water Reactor

By making use of the isotope approximation; neglecting squares of relative mass differences among the isotopes, the authors derived analytically approximations to ordinary diffusion coefficients in a 3- and 4- component isotope mixture. Moreover, approximations to multi-component diffusion coefficients were given on the analogy of those to the 3- and 4-component coefficients, and these approximations were verified to satisfy constrains on the exact ordinary diffusion coefficients.

For 4-component mixture of uranium hexafluoride isotopes, ²³⁴UF_6- ²³⁵UF_6- ²³⁶UF_6- ²³⁸UF₆, composition dependences of the approximation were equal to those of the exact diffusion coefficients. In addition, relative errors between the exact and the approximations were less than 0.2% for 5-component mixture of krypton 80–82-83-84-86 isotopes.     

压水堆棒控系统数字化设计

系统采用西门子S7-400 PLC进行系统的总体控制,采用PROFIBUS-DP网络进行系统的通讯,实现了棒控系统的数字化控制,并对棒控系统的硬件设计、软件设计以及系统的调试进行了研究。该系统设计新颖,技术先进,符合国际发展趋势,对国产核电站数字化仪控系统的出口具有现实意义。     

超临界水冷堆开发现状与前景展望

超临界水冷堆是被国际上选定为第四代核能系统长远开发的6种堆型之一,是在现有LWR和超临界火电技术基础上发展起来的革新型设计.在技术上,超临界水冷堆可以借鉴现有PWR和超临界火电的设计、建造和运行经验,不存在不可逾越的技术障碍.我国近期和中期目标都是采用压水堆技术,考虑到技术的继承性和可持续发展的要求,开发和研制超临界水冷堆核能系统是必然的选择.     

反应堆退役废金属高压水去污废水的再利用

用高压水射流去污将反应堆退役废金属再循环再利用,会产生相当数量的放射性废水。为了节省放射性废水处理费用并减少清水的消耗量,降低废金属去污的成本,应将这部分放射性废水回用。高压水去污废水中的放射性主要来自其中的固体颗粒。因此回用废水就要将水中固体颗粒去除。研究分析了各种可能的工艺,经过代价利益分析,选用了离心分离工艺。     

压力容器水位参数在堆芯损伤评价方法中的应用

事故工况下,堆芯会随着冷却能力的下降而逐步升温,长时间的裸露会导致堆芯损伤,而堆芯出口温度和压力容器水位可直观反映堆芯的冷却能力。以西屋公司堆芯损伤评价导则为基础的堆芯损伤评价方法将堆芯出口温度和安全壳剂量率作为主要参数评价堆芯损伤状态,压力容器水位作为辅助参数之一来验证评价结果的合理性,但一些核电厂堆芯出口热电偶量程并不能满足严重事故条件下的要求,需要其他替代参数。本工作以压水堆核电厂严重事故分析数据为基础,探讨将压力容器水位作为主要参数应用于堆芯损伤评价方法的可行性。     

高压水清洗技术在反应堆退役工程中的验证试验

在反应堆退役工程中使用高压水清洗技术对反应堆工艺运输水池、工艺房间、密闭水池和一些特殊设备进行了清洗去污。通过施工前的工程验证试验及工程施工，取得了高压水清洗技术应用于大型核设施去污的1套技术数据。同时针对特殊情况进行了技术改进，使该技术在核设施退役领域中得到进一步完善。     

Experimental Study on Reactivity Worth for Absorber Material in High Conversion Light Water Reactor Using FCA-HCLWR Core Fueled with Enriched Uranium

Experimental study on reactivity worth for absorber material in HCLWR core has been carried out in a series of experiments using the Fast Critical Assembly (FCA) in Japan Atomic Energy Research Institute (JAERI). The central reactivity worth as well as the simulated control rod worth of B₄C with different ¹⁰B content and of Hf was measured in FCA-HCLWR core fueled with enriched uranium. Both reactivity worths of B₄C increase with ¹⁰B content. These increasing trends do not saturate to 90% enriched B₄C. The Hf has the smaller reactivity worth than the 20% B₄C. The experimental values are compared with the calculated ones which obtained from JENDL-2 data and the SRAC system. The calculation predicts well the dependence of reactivity worth on ¹⁰B content and underestimates the reactivity worth ratios of the Hf to the 20% B₄C.     

Friction Pressure Drop and Heat Transfer Coefficient of Two-Phase Flow in Helically Coiled Tube Once-Through Steam Generator for Integrated Type Marine Water Reactor

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report.     

高压及超临界压力下螺旋管内单相水摩擦压降特性实验研究

在较宽的参数范围内对高压及超临界压力下单相水流经立式螺旋管的摩擦压降特性进行实验研究。研究表明,对于高压工况,在高雷诺数Re条件下的水力粗糙区,管壁粗糙度对摩擦压降的影响要远大于二次流的影响,故单相水在螺旋管内的摩擦压降基本可直接采用直管的相关公式来预测;对于超临界压力工况,在物性变化剧烈的拟临界区内,采用高压工况计算公式所得的摩擦压降预测值要明显偏低于实验值。     

多用途一体化轻水堆初步设计方案和安全分析

为满足未来区域性核能供电、核供热、大规模制氢、海水淡化等需求,迫切需要一种结构简单、固有安全性高、经济性高的多用途反应堆.基于此,一种多用途的一体化轻水堆设计概念被提出,包括不同设备的初步设计方案和参数;根据其特点,利用最佳估算程序RELAP5对其中一个设计方案进行了稳压器汽腔破口事故和主泵断电引起的丧失流量事故的确定论安全分析.结果表明,在保守假设条件下,其固有特性和安全系统仍能保证堆芯始终处于被淹没状态,非能动余热排出系统可有效导出停堆后的长期衰变热,从而为进一步研究一体化轻水堆的设计和运行安全特性打下了基础.     

反应堆控制棒新型电磁驱动机构原理实验监控系统

为解决反应堆控制棒新型电磁驱动机构工程样机设计中遇到的一些问题，对该机构的原理样机进行了性能实验。该原理样机实验监控系统采用光栅尺测量控制棒的实际位移，可编程控制器(PLC)控制实验台，计算机采集实验数据，计算机与PLC之间通过RS232串口进行通信。实验表明：该实验监控系统性能可靠，精度高，保证了实验的顺利完成。     

轻水堆严重事故及可能的缓解措施

现有概率安全评价指出，常规轻水堆的堆芯熔化频率及安全壳失效，放射性大量释放的频率都是是很低的。但这些风险对于下一代先进轻水堆说是不能忽力听，近年来西方对下一代先进轻水堆的安全目标作了更高的要求，即在严重事故的条件下，仍然能保证安全壳的完整性，而无需采取应急措施，这就要求对严重事故现象可有足够的认识，以便对严重事故设置相应的缓解措施，本文简述了严重事故的物理现象，机理及可能的缓解策略，综述了这方面的     

压水堆棒控系统动力熔断器烧毁原因分析及维修策略改进

对某核电厂反应堆棒控系统动力熔断器烧毁事件进行根本原因分析。通过对反应堆棒控系统原理和熔断器所在电路的分析,结合熔断器熔断特性和可控硅导通与关断特性,实测数据验证动力熔断器烧毁根本原因是可控硅性能异常导致相邻两相电压短路,相间短路造成两相动力熔断器瞬间熔断。并给出反应堆棒控系统维修策略改进建议。     

水平光管内超临界水在大比热容区内的传热与交混特性数值模拟研究

对超临界压力水在管径为32 mm 3 mm、长度为8 m的水平光管内的流动和传热特性进行数值模拟研究。探讨不同压力、流量、热负荷下管内换热系数的变化特征;重点分析超临界水的交混特性,对比分析流动通道内二次流动的特性及演变规律,进而对二次流动的变化规律给出合理的解释;利用无量纲的Gr/Re2和Gr/Re2.7对交混特性中自然对流与强制对流的相对大小进行定量描述,以解释超临界水在水平管内的流动与换热特征。     

汽轮机调节系统二次脉动油压异常分析

针对田湾核电站汽轮机夏季运行工况时出现调节系统二次脉动油压高且波动问题,采用两台机组汽轮机调节系统静态试验曲线以及热力系统参数对比分析的方法,阐述了循环冷却水温度、二次脉动油压拐点、调节阀空行程对二次脉动油压的影响。根据研究结果,通过提高二次脉动油压拐点及减小高压调节阀油动机空行程的纠正措施,有效解决了此问题,提高了调节系统的稳定性。对机组的运行和维修具有一定的借鉴意义。     

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

Effect of Microstructure on Failure Behavior of Light Water Reactor Coolant Piping under Severe Accident Conditions

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e., in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% proof stress.     

ATWS Characteristics of Super LWR with/without Alternative Action

Anticipated-transient-without-scram (ATWS) of the supercritical-pressure light water cooled thermal reactor with downward-flow water rods (Super LWR) is analyzed to clarify its safety characteristics. At loss-of-flow, heat-up of the fuel cladding is mitigated by the water rods removing heat from the fuel channels by heat conduction and supplying their coolant inventory to the fuel channels by volume expansion. The average coolant density is not sensitive to the pressure due to the small density difference between “steam” and “water” at supercritical-pressure. Closure of the coolant outlet of the once-through system causes flow stagnation that suppresses an increase in the coolant density due to an increase in the temperature. Therefore, the increase in power is small for pressurization events. The coolant density and Doppler feedbacks provide good self-controllability of the power against loss-of-flow and reactivity insertion. An alternative action is not needed either to satisfy the safety criteria or to achieve a high-temperature stable condition for all ATWS events. Initiating the automatic depressurization system is a good alternative action that induces a strong core coolant flow and inserts a negative reactivity. It provides an additional safety margin for the ATWS events. Even the high core power rating of the Super LWR has excellent ATWS characteristics, providing a key reactor design advantage.