The Fort St. Vrain high temperature gas-cooled reactor: V. Steam system performance tests

The Fort St. Vrain primary and secondary coolant systems have given satisfactory performance during the rise to power test program with the tests being terminated at the current maximum allowable thermal reactor power of 70% of rated. Because of a regenerative heat problem in the steam generators, rated conditions of 1000°F main and hot reheat steam temperatures, predicted to occur at 25% power, were not reached until 68%. The regenerative heat problem also forced “overblowing” of the core with primary coolant helum which resulted in higher fuel temperatures than predicted, lower core primary coolant outlet temperatures and higher core primary coolant inlet temperatures. Data suggest that all parameters will be at rated conditions at 80–100% power. A small steam generator tubing leak was detected by the primary coolant moisture monitors of the plant protective system. It was located by covergas techniques and repaired by plugging the leaking feedwater and steam subheaders external to the reactor.     

The design of 2300 MW(e) twin high temperature gas-cooled reactors for the philadelphia electric company

The basic design features of a 2300 MW(e) twin high temperature gas-cooled reactor (HTGR) power plant are described. The reactor core consists of vertical columns of hexagonal graphite fuel-moderator elements and graphite reflector blocks which are grouped into a cylindrical array and supported by a graphite core support structure. Reactivity control is accomplished by means of 146 control rods. The distribution of helium coolant flow through the core is controlled by variable orifice valves. Each of the six primary coolant loops is equipped with a helium circulator. The main steam/water section of each steam generator consists of a single helical tube bundle arranged in an annulus around the center duct. A core auxiliary cooling system is provided to furnish an independent means of removing reactor afterheat. The inherent safety characteristics and the design safety features of the large HTGR are discussed. Station arrangement, steam cycle and twin turbine generators, plant performance and control, containment and fuel handling, and environmental controls, are described.     

The fort St. Vrain high temperature gas-cooled reactor: IX. Rise-to-power physics tests

Predicted and measured radial region peaking factors differed more than anticipated, most especially in core regions that were susceptible to ingress of diverted cool primary coolant helium into the thermocouple tubes where the individual region outlet temperatures were measured. To compensate for such cooling, inferred region outlet temperatures were empirically determined by parameter comparison to similar regions in the core not subject to ingress of diverted, cool primary coolant. Decalibration (loss of linearity in indicated versus actual reactor power) existed for both the source range (startup) neutron detectors and the power range neutron detectors. In both cases operational restraints were established to compensate for the lack of linearity.     

The Fort St. Vrain high temperature gas-cooled reactor: XII. The dew point moisture monitor testing program

Moisture ingress into the core volume could cause damaging reactions with the moderator-reflector graphite and burnable poison, therefore a dew point moisture monitoring system has been developed with the basic design criteria that a plant protective system trip is signaled after the system detects high primary coolant helium moisture levels and that the system is able to correctly identify which of two steam generator loops is leaking. Modifications to the sample supplies to the monitors were necessary to reduce the system's unsatisfactory response time at lower reactor power levels.     

船用反应堆冷却剂泵连续调速研究

采用船用核动力装置模拟程序,对反应堆冷却剂泵转速连续调节研究进行仿真试验研究。在相同的40%满功率工况下,进行冷却剂泵转速阶跃变化与连续变化两种试验。对比了反应堆进出口温度、反应堆功率、反应堆反应性、冷却剂流量、蒸发器蒸汽压力等参数的变化情况,对开展船用反应堆冷却剂泵连续调速设计具有重要的指导意义。     

10MW高温气冷堆控制方案研究

在清华大学核能设计研究院开发的高温堆可视化仿真控制平台上进行了10MW高温气冷堆动态特性研究,并结合其运行特点和控制要求设计了3种控制方案,采用比例积分与微分控制方法,在高温堆可视化仿真控制平台上进行了控制方案的仿真比较。控制的重点在于维持直流蒸汽发生器的出口蒸汽温度恒定,同时兼顾反应堆出口热氦气温度不超出保护限值。仿真结果表明,采用给水泵调节给水流量来控制蒸汽温度,并通过氦风机调节氦流量保持与给定功率成比例,避免跨回路调节,静态解除了由于氦流量的变化对一、二回路的耦合问题,能够获得理想的控制效果。     

运行条件对蒸汽发生器热工参数影响的仿真研究

以减轻蒸汽发生器破管事故及考察核电站电力升级为目的,参考大亚湾核电站蒸汽发生器的运行参数,基于分布参数法建立了核动力蒸汽发生器一维数学模型,开发了基于MATLAB的动态仿真程序,进行了改变运行条件时蒸汽发生器热工参数仿真计算。计算结果表明:与满负荷正常运行条件相比,在降低二回路运行温度或增加二回路流量时,二回路预热段变短,出口焓大幅升高;质量含汽率在降低温度时提高54%,增加流量时提高28%;一、二回路及管壁整体温度降低;一回路和内壁温降增大。该计算结果揭示了蒸汽发生器的内在传热规律,可为缓解U形管恶化及提升电力的相关操作提供一定理论依据。     

Measurement of subassembly outlet coolant temperature in the JOYO experimental fast reactor

Thermocouple temperature sensors are installed above the central region of the core in the JOYO experimental fast reactor to monitor the outlet coolant temperature of 115 subassemblies. This paper summarizes the experimental temperature data obtained during initial 50 MWt operation of the reactor. Subassembly outlet coolant temperature distributions that were obtained under various power levels, different main cooling system flowrates, and unequal reactor inlet temperatures from the two cooling loops are described. In addition, coolant temperature and flowrate distributions at the subassembly outlet measured in a zero power experiment are presented.     

全厂断电事故下AP1000非能动余热排出系统分析

利用RELAP5/MOD3.3程序对AP1000反应堆一回路及非能动系统进行建模计算,给出了AP1000非能动余热排出系统（PRHRS）在全厂断电事故下的瞬态响应特性。计算结果表明：情况1,PHRH系统由蒸汽发生器低水位与低启动给水流量符合信号启动,稳压器安全阀的开启导致PRHRS发生倒流现象,并会引起堆芯冷却剂过热沸腾、压力容器进出口温差过大等后果;情况2,由断电信号直接触发PRHRS,触发前安全阀不开启,此时PRHRS正常运行。     

不同运行工况下蒸汽发生器热工水力稳态特性数值研究

基于相似模化理论建立了蒸汽发生器一、二回路流体及传热管流 固耦合传热的单元管三维物理模型,对大亚湾核电厂蒸汽发生器不同工况下的热工水力稳态特性进行了数值模拟研究。采用热相变模型描述二回路汽液两相流动与换热、流-固耦合模型描述一回路冷却剂借助U型管与二回路流体换热。数值计算结果表明：满负荷运行时,传热管内壁温度变化趋势与一次侧流体基本一致,外壁温度与二次侧流体温度变化趋势相同;截面平均含汽率沿传热管高度的升高呈上升趋势,出口质量含汽率与大亚湾核电厂实际运行参数相符;随负荷降低一回路出口温度基本不变,二回路出口温度升高,质量含汽率及传热系数下降,平均传热系数与Rohsenow经验关联式的计算结果基本吻合。     

Simulating test for thermal mixing in the hot gas chamber of the HTR-10

The helium coolant at the outlet of the pebble bed core of the 10 MW High Temperature Gas-cooled Reactor-Test Module exhibits a severe radial temperature deviation. In order to avoid damages at the downstream components due to alternating thermal loads such as the steam generator, a hot gas chamber is especially designed to solve the problem. Thermal mixing performance of the coolant in the hot gas chamber is experimentally investigated on a 1:1.5 scale model by air. The experimental result shows that within the Reynolds number range of 1.4×10⁵–5.8×10⁵, the hot gas chamber with a radial mixer reaches excellent thermal mixing of the coolant of about 94%. The flow resistance coefficient for the hot gas chamber is also presented.     

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.     

蒸汽发生器工作过程建模及仿真分析

基于分布参数热工对象的集总参数化动力学模型,对自然循环蒸汽发生器进行了控制体划分并建立了数学模型,并用MATLAB语言和SIMULINK仿真软件对其进行了仿真,文章采用了Runge-Kutta (4,5)求解器,得到不同功率装置运行时,一次侧,二次侧,以及管束的温度分布,并得到一回路给水扰动时,传热量以及冷却剂出口焓的响应曲线.     

The Fort St. Vrain high temperature gas-cooled reactor: XI. Control rod drive and orifice assemblies

Cable-actuated control rods driven by an electrically powered winch system have performed successfully during preliminary operating tests and during the rise to power test program. Initial fitting and other problems were encountered which were largely associated with excessive moisture in the normally dry primary coolant helium.     

The choice of optimum parameters for atomic power stations using a gaseous coolant

The article discusses a procedure for choosing the optimum parameters for atomic power stations with gaseous cooling of the reactors. These parameters correspond to a minimum value of the calculated cost of the power station. The methods take into account the thermal and economic characteristics of the power station, as well as the specific limitations of the maximum permissible surface temperature of the fuel elements. The article considers conditions where the thermal power of the reactor is given, as well as the more general case in which the optimum power of the reactor is to be determined together with the other desired power station parameters.Analytic design formulas are given for an approximate determination of the optimum values of coolant temperature rise in the reactor and of coolant temperature at the outlet of the steam generator.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Design and development of a steam turbine driven circulator for high temperature gas-cooled reactors

Design optimization of high temperature gas-cooled power reactors with integration of the entire primary coolant system in a prestressed concrete reactor pressure vessel (PCRV) leads to stringent power, size, and environmental requirements for the primary coolant circulation equipment. To meet these requirements, the series steam turbine helium circulator (SSTHC) was conceived and developed by Gulf General Atomic. The SSTHC consists of a single stage axial-flow helium compressor driven by a single stage steam turbine. The compressor and the drive turbine are integrally attached to a single vertical shaft and overhung from a central bearing and seal housing. The drive turbines are in series with the power-producing main steam turbine.This paper discusses the design points of the compressor, drive turbine and auxiliary Pelton wheel drive, as well as the design requirements for the bearings and seal system. A general outline of the SSTHC development program carried out at Gulf General Atomic is given. The following areas are included in the development program: aerodynamics, compressor noise, primary coolant shutoff valve, water bearings and rotor dynamics, seals, blade vibration, and disk catcher. Further, a comprehensive series of transient tests on a circulator have been carried out.     

Analysis of fuel element matrix graphite corrosion in HTR-PM for normal operating conditions

The corrosion of spherical fuel element by oxidizing gases will degrade their thermal and mechanical properties in pebble-bed reactors. Oxidizing impurities in primary helium coolant may have significant influence on the gasification of graphite during the long service time even though their concentrations remain relatively low. This paper deals with the corrosion of matrix graphite by steam and oxygen in Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM). The influence of steam contents was first analyzed, and then the effect of oxygen contents was also taken into account. The results show that the corrosion by steam was weak and it would not threaten the normal service of spherical fuel elements for expected steam contamination levels. On the contrary, the corrosion would be more severe while the oxygen content was as high as currently expected. Finally, the upper limits of steam and oxygen in primary helium coolant were recommended to be 1.0 and 0.05 cm³ m⁻³.     

The Fort St. Vrain high temperature gas-cooled reactor: II. Helium circulators

In field tests in a fossil-fueled facility, performed concurrently with Fort St. Vrain's construction, data indicated that the helium circulator design was well suited to provide primary coolant circulation for the high temperature gas-cooled reactor. After plant installation, primarily during the hot functional tests, a number of time-consuming delays developed caused by cavitation damage on circulator speed valves, cavitation and fatigue damage on auxiliary water turbine buckets, water turbine nozzle erosion, static shutdown seal cracks and circulator primary closure helium leakage. After extensive analysis and testing, all of these problems were corrected. Circulators have performed satisfactorily at levels up to 70% of rated power.     

Preliminary issues associated with the next generation nuclear plant intermediate heat exchanger design

The Next Generation Nuclear Plant, with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 850-950 °C. In this concept, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, a nitrogen/helium mixture, or a molten salt. This paper assesses the issues pertaining to shell-and-tube and compact heat exchangers. A detailed thermal-hydraulic analysis was performed to calculate heat transfer, temperature distribution, and pressure drop inside both printed circuit and shell-and-tube heat exchangers. The analysis included evaluation of the role of key process parameters, geometrical factors in heat exchanger designs, and material properties of structural alloys. Calculations were performed for helium-to-helium, helium-to-helium/nitrogen, and helium-to-salt heat exchangers.