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.     

超临界水堆控制条件下扰动特性分析

超临界水堆的一次通过循环设计不同于现有轻水堆,因此研究其扰动特性十分重要。在发生扰动后欲保持电站运行稳定,就要依靠控制系统调节达到稳定的状态。本文通过FORTRAN编制程序,研究以控制棒、汽轮机控制阀与反应堆冷却剂泵为控制方式的电站系统中,发生压力、温度等扰动时,反应堆内参数的变化。结果表明：给水流量的扰动不会对系统行为有很大影响,给水温度下降的扰动需较长时间才能达到稳定,压力设定值变化扰动时稳定所需的时间较短。     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.     

Supercritical-pressure, light-water-cooled reactors for economical nuclear power plants

Design studies of supercritical-pressure light-water-cooled reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is cooled by supercritical water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedwater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.     

A Linear Stability Analysis of Supercritical Water Reactors, (I)

This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient at the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability.     

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.     

超临界水冷堆的安全分析

应用RELAP5-3D程序建立了超临界水冷堆(SCWR)的稳态模型,并在此基础上,分别对SCWR的两种瞬态和两种事故工况进行了分析。汽轮机旁路系统的存在可有效维持反应堆压力,保证反应堆安全。若SCWR失去给水,在辅助给水系统启动之前,向下流的水棒可通过热传导带走堆芯热量,并向燃料通道内提供冷却剂,缓解堆芯升温。因而,向下流的水棒体现了SCWR的安全性。主泵卡轴事故由于没有惰转,最热包壳温度值最大,因而主泵惰转可有效缓解包壳温度的升高。     

压水堆一回路卸压措施影响因素研究

研究压水堆一回路管道小小破口失水事故叠加辅助给水失效导致的高压堆芯熔化严重事故进程，对比验证不同严重事故缓解措施入口温度条件下一回路卸压缓解途径的充分性和有效性，并确认较佳的一回路冷却系统（RCS）降压途径。结果显示，以低于650℃的温度作为降压缓解措施入口条件，可及时恢复可能的堆芯冷却能力。一、二回路卸压效果分析表明，考虑了长期衰变热移出注水流量和堆芯过冷度要求，较佳的卸压配置为初期打开一列稳压器卸压阀，同时迅速恢复辅助给水并开启蒸汽发生器卸压阀。      

Development and diversity and defense-in-depth application of ABWR feedwater pump and controller model

This work developed an advanced boiling water reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element/three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including step change of dome pressure, feedwater pumps transfer, inadvertent closure of all turbine control valves, and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of stand-alone model and incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an emergency core cooling system (ECCS) automatic diverse back-up. In Lungmen nuclear power plant (NPP), a diverse manual initiation means for the high pressure core flooder (HPCF) loop C is designed as the back-up of digitalized engineered safety features actuation system (ESFAS). If the motor-driven feedwater pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common-cause failure (CCF).     

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.     

Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.     

压水堆核电厂主给水管道水锤计算及分析

核电厂二回路主给水系统是保证蒸汽发生器冷却的重要系统,同时也是水锤频发的管段,研究水锤对主给水系统的规律对于系统稳定运行具有重要意义。文章以主给水系统作为研究对象,通过Flowmaster软件的瞬态计算功能,建立给水泵、控制阀门等边界条件下的数学模型,计算阀门、泵关闭时产生的水锤压力,并且导出压力等参数的瞬时变化解。结论验证了Flowmaster瞬态计算功能计算水锤的可行性,结合工程实例说明,增加给水控制阀、给水泵关闭时间能有效控制水锤,改变给水泵、给水控制阀关闭触发信号间隔能缓解水力冲击,以及管径等因素对水锤的影响,对于实际工程中的设计和系统优化具有一定指导作用。     

超临界快堆给水控制失效瞬态控制分析

超临界快堆是一次通过循环,瞬态安全特性不同于现有的轻水堆.以控制棒、汽轮机主进汽阀、反应堆冷却剂泵作为超临界快堆的控制方式,在给水控制系统失效瞬态事故工况下,研究该堆采用不同控制方式时,反应堆内压力、功率、冷却剂温度、冷却剂质量流量及包壳表面温度等参数随时间的变化情况.结果表明:采用汽轮机主进汽阀与控制棒联合控制时,反...     

超临界水冷堆述评

超临界水冷堆(SCWR)是在高于水的临界点（374℃，22.1MPa）的温度和压力下运行的反应堆。相对于传统的轻水堆，它的热效率显著提高，可达45％。由于冷却剂在超临界状态下不发生相变，可直接与能量转换设备相联，从而简化了反应堆的结构。在SCWR中不需再循环和射流泵、稳压器、蒸汽发生器、汽水分离器和干燥器。它的主要特点是经济性好。     

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

A Concept of Passive Safety,Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered.

The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core.

In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of “safety class” for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system.

The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

中国一体化反应堆核电厂创新安全壳设计研究

中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析.