Influence of the reactor temperature characteristics upon the choice of the optimum thermodynamic cycle of an atomic-electrical generating station

In this article we investigate the possible temperature changes in a nuclear energy station as a function of the thermal power of the reactor when there are two limiting temperatures: for the shell and for the center of the heat-emitting elements. We find the changes in the allowed thermal and electrical power of the unit as a function of the parameters of the thermodynamic cycle. We give the reader an understanding of the boundary thermal power of the reactor and the efficiency of the generator.We also give the conditions under which the formulas which we have derived may be used for a preliminary calculation of the optimum parameters of the thermodynamic cycle. Our analysis gives the curves which show the increase in the parameters and the efficiency of the electrical generating station as a function of the material of which the shell is constructed and of the type of nuclear fuel used.     

Selection of coolant parameters for removal and utilization of heat from reactors

The conditions for choosing inlet and exit reactor coolant temperatures are discussed. The reactor exit temperature, for both power and dual purpose reactors, is determined by the allowable temperature for the cladding material of the fuel elements. The coolant inlet temperature in a power reactor is determined by the produced steam parameters, and in a dual purpose reactor it is determined by the condenser cooling water temperature or the melting temperature of the coolant when liquid metals are used. In connection with this, the thermal efficiency, which is always greater in power reactors than in dual purpose reactors, is discussed. In spite of the lower thermal efficiency in dual purpose reactors, the total of energy produced is but little different from the energy produced in power reactors. However, if power reactors are designed for the maximum limits of heat flux, they become more profitable than dual purpose reactors since at the same thermal power level they allow generation of a greater amount of electrical energy.     

Selection of coolant parameters for removal and utilization of heat from reactors

The conditions for choosing inlet and exit reactor coolant temperatures are discussed. The reactor exit temperature, for both power and dual purpose reactors, is determined by the allowable temperature for the cladding material of the fuel elements. The coolant inlet temperature in a power reactor is determined by the produced steam parameters, and in a dual purpose reactor it is determined by the condenser cooling water temperature or the melting temperature of the coolant when liquid metals are used. In connection with this, the thermal efficiency, which is always greater in power reactors than in dual purpose reactors, is discussed. In spite of the lower thermal efficiency in dual purpose reactors, the total of energy produced is but little different from the energy produced in power reactors. However, if power reactors are designed for the maximum limits of heat flux, they become more profitable than dual purpose reactors since at the same thermal power level they allow generation of a greater amount of electrical energy.     

船用压水堆核动力装置双恒定运行方案静态特性研究

讨论了船用压水堆核动力装置的双恒定运行方案以及实现的技术手段 ,并通过反应堆热工安全准则的计算和蒸汽发生器传热实验 ,从稳态运行过程的角度探讨了船用核动力装置实现双恒定运行方案的可行性。     

超高通量快中子研究堆需求分析

实现超高快中子通量是世界先进研究堆的重要发展方向,对于加快第四代先进核能系统燃料及材料创新发展具有重要意义。本文从先进核能堆内结构材料与核燃料的辐照考验、长反应链超钚元素生产等角度,初步分析了我国建设超高通量快中子研究堆的必要性。在此基础上,确定了超高通量快中子研究堆的堆芯最大中子注量率及其冷却剂,给出了反应堆主要参数及冷却剂流动方案。反应堆热功率为200 MW,冷却剂为铅铋合金,最大中子注量率大于1016 cm?2·s?1。      

The optimum design of tight lattice reactor core with thin rod bundles

The fuel height, rod diameter, pitch, and the loading pattern are all important parameters in the reactor core design process. Based on the analysis of the core performance, optimization calculation is performed on the three objective functions of ABV-6M reactor, i.e., power density, coolant temperature difference between the inlet and outlet, and flow-induced vibration are proposed for optimization calculation. Then a multi-objective problem (MOP) model is applied and computed optimally by non-dominated sorting genetic algorithm (NSGA-II) with the aim of maximizing power density and temperature difference as well as minimizing the flow-induced vibration. The results of optimal designs called ‘Pareto-optimal solutions’ are a set of multiple optimum solutions, from which the final optimization can be chosen after sensitivity analysis is performed. On the basis of lattice parameters optimization, the radial one-dimensional fuel loading pattern was optimized for achieving the optimum fuel utilization. The typical optimum design considered to be safe in a verification check showed that tight lattice effectively improved the reactor performances and saved the fuel consumption.     

Thermo-hydraulic analysis for SCWR during power-raising phase of startup

The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR.     

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.     

On some methods for raising the power level of reactors with gaseous coolants

The present article is concerned with certain methods for raising the power level of reactors with gaseous coolants: additional cooling of the gas ahead of the gas compressor, increasing the pressure in the loop, and profiling the coolant flow through the reactor.Equations for calculating the theoretical thermodynamic cycle for a reactor with a gaseous coolant, the coolant temperature at the downstream end of the reactor, and the profiling of the coolant flow are derived.     

The feasibility of cooling heavy-water reactors with supercritical fluids

It appears technically feasible to use supercritical carbon dioxide as a coolant for a CANDU-type reactor. A new supercritical loop is proposed in which the reactor is cooled by a single-phase fluid pumped in a high density liquid-like state. The supercritical fluid-cooled reactor has the advantage of gas-cooled reactors of avoiding dryout, and of liquid-cooled reactors of low coolant-circulation power. By eliminating dryout, the maximum operating temperature of the fuel sheath can be increased to 1021°F (550°C) for existing Canadian fuel bundles, with a coolant exit temperature of 855°F (458°C) producing steam comparable to that of conventional fossil-fuel plants. Since the reactor coolant exit temperature from the steam generator may be as high as 280°F (138°C) low-pressure steam may also be produced. A new dual-reheat cycle is proposed with an ideal overall plant efficiency of 33%, comparable to the Pickering generating station.     

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.     

双环路压水堆非对称入口条件下物理-热工特性研究

双环路压水堆存在反应堆入口流量、温度不对称的非正常运行工况。本文建立了基于CFD方法的反应堆整体三维流场模型,并耦合中子动力学计算程序和RELAP5程序,对这种非对称入口条件下的反应堆物理-热工特性进行了数值模拟。结果表明：反应堆入口流量不对称会加剧堆芯入口流量分配的不均匀性,并进一步导致局部功率变化,对反应堆安全不利;在入口温度不对称的条件下,冷却剂在下腔室的混合非常不充分,并导致堆芯入口温度分布不均匀,引起局部功率变化较大,对反应堆安全不利。     

The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor

In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (T_in) and the temperature difference between the coolant inlet and outlet (ΔT) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong dependence of the reactor power on coolant temperature in agreement with the design of MNSR. The experimental parameters were found to be in good agreement with data obtained using a semi-empirical relationship between the reactor power, flux parameters, core inlet temperature, and the coolant temperature rise. The relationship was therefore used to predict the power level of NIRR-1 from its neutron flux parameters to which it has been found to be proportional. The variation of T_in and ΔT with the reactor power and flux was also investigated and the results obtained are hereby discussed.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Core design of a high-temperature fast reactor cooled by supercritical light water

A high-temperature large fast reactor cooled by supercritical water (SCFR-H) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. We study the reactors with blankets cooled by ascending and descending flow. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low, 467°C. The reasons are as follows: (1) the power swing due to the accumulation of fissile material in the inner blankets with burn-up, and (2) local power peak in the assemblies due to the zirconium-hydride layers. The difference in the outlet coolant temperature is more enhanced than the low temperature core where outlet temperature is approximately 400°C. The reason is that the coolant temperature is more sensitive to the enthalpy change than near the pseudo critical temperature, 385°C at 25 MPa. Thus, we design the core with blankets cooled by descending flow to obtain high coolant outlet temperature. The coolant outlet temperature becomes 537°C, which is 70°C higher than that of the core with ascending blanket flow. The thermal efficiency is improved from 43.2 to 44.6%. The coolant mass flow rate per electric power decreases by 14%. This will reduce the size of the balance of plant (BOP) system. The power of the reactor is high (1565 MWe) and the void reactivity is negative.     

高温气冷堆热工分析模型改进与堆芯温度场分析

更准确地模拟球床式高温气冷堆堆芯温度分布,是反应堆安全分析尤其是超高温运行研究中的关键问题之一。由于堆芯球流运动具有不确定性,石墨块和碳砖等结构材料采用散体布置,堆内冷却剂流道复杂,对热工水力准确模拟造成困难,可进一步优化。本文结合HTR 10的结构特点和流道特征,简要分析了堆芯传热过程,说明了在热工模拟中准确划分结构和流道对获取更精确的堆芯温度分布的重要意义。详细梳理了冷却剂流动路径,改进了在THERMIX程序下建立的HTR 10原有热工分析模型,更合理地模拟了堆芯冷却剂漏流行为,使得模型对堆芯冷却剂流动和传热过程的描述更准确。与试验数据对比,改进后的模型对堆芯外围系统的温度分布模拟准确性显著提升。计算结果表明,反应堆在额定设计工况下满功率稳态运行时,燃料和反射层最高温度均未超过材料的耐热限值。     

棱柱式高温气冷空间核反应堆初步方案设计与中子物理分析

空间核反应堆电源将核裂变能转换为电能,与太阳能、化学燃料电池等其他形式的电源相比,具有电功率大、系统比功率高、使用寿命长等优点,在太空探索中具有广阔的应用前景。以高温气冷堆技术为基础,提出了以氦氙混合气体作冷却剂,直接布雷顿循环的空间核反应堆电源方案。核反应堆是采用包覆颗粒燃料的小型棱柱式高温气冷堆,热功率为5 MW。采用蒙特卡罗方法进行了中子物理分析。结果表明,设计的反应堆满足10a以上的满功率运行寿期,具有负的反应性温度系数。通过布置B4C安全棒,使反应堆在发射失败引起的堆芯进水事故中能保证次临界。     

核电站反应堆冷却剂系统主管道安装焊接

以秦山核电二期工程为例,论述了核电站反应堆冷却剂系统主管道安装焊接技术及质量控制要点,并对反应堆冷却剂系统主管道的安装顺序、安装技术要求、焊接质量检验方法以及焊接变形的控制等方面给予了详细的阐述,对核电站反应堆冷却剂系统主管道安装焊接及质量控制具有借鉴作用。     

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.     

一体化压水堆强迫循环转自然循环过渡过程特性分析

针对一体化压水堆核动力装置,以核动力装置瞬态最佳估算程序RELAP5/MOD3为基础,采用两群三维时空中子动力学模型替代点堆模型,并建立三维空间内中子物理与热工水力的耦合模型,研制相应的计算程序。对一体化核动力装置强迫循环向自然循环转换过程进行仿真模拟。在过渡过程中,一体化压水堆核动力装置反应堆功率变化幅度较大,冷却剂流量的变化对一回路温度影响较大。