Innovative nuclear technology based on modular multi-purpose lead–bismuth cooled fast reactors

Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development.

One of the ways of finding the solution to the problem can be the use of modular fast reactors SVBR-75/100 cooled by lead–bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines.

The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors (100 MWe), chemical inertness and high boiling point of lead–bismuth coolant, integral design of the pool type primary circuit equipment.

Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway.

Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper.     

Development of medium and small sized reactors: DMS

Nuclear power is expected to become the main source for electric power generation in Japan for the reasons of energy security and prevention of CO₂ emission. In addition, the slowdown of recent electric power demand and the liberalization of the electric power market are accelerating medium and small sized reactor development. (Hida and Ito, 2003) Furthermore, the needs of medium and small sized reactors have become greater in foreign countries where electric grid systems are weak. Under these circumstances, Hitachi has developed DMS's (Double MS: Modular Simplified & Medium Small Reactors) as 400 MWe class LWR's supported by The Japan Atomic Power Company. (Moriya et al., 2003) In addition, DMS's have been designed based on proven technology that requires no large-scale development, and can therefore be introduced in the market in near future.     

A rail transportable lead–bismuth cooled reactor

A conceptual design of a small reactor cooled by lead–bismuth is developed. The main constraint on this reactor design is its transportability. The whole reactor module should be transportable on a rail cart. This imposes a volume envelope of approximately 4.5 × 4.5 × 24 m and the maximum weight of about 300 tons. Therefore, the reactor vessel is 3 m in diameter and 3.85 m tall. In order to satisfy the proliferation resistance requirements the reactor is sealed after the fuel is loaded and shall not be opened until it is shipped back after it reaches its end of lifetime after 15years. The reactor fuel is 11% and 13% enriched plutonium nitride. Reactor power is 50 MW_th which translates into 15 MW_e. Reactor pool is at nearly atmospheric pressure. Core inlet and outlet temperature are 350 and 365 °C, respectively. The reactor uses electromagnetic pumps to drive the primary coolant circulation. Secondary system consists of saturated steam cycle operating at 7 MPa and 290 °C. This reactor is well suited for secluded areas with the demand for electricity such as small islands.     

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.     

Modeling and optimization of steam dryer for removal of lead–bismuth droplets

Removal of lead–bismuth droplets from steam flow is a crucial issue in the direct contact boiling lead–bismuth cooled fast reactor. Droplets are generated due to the boiling of water directly in the reactor chimney, where steam for the turbine is generated. The droplets could severely damage the turbine and therefore a steam dryer is used for their removal. This paper presents an optimization of the main steam dryer geometrical parameters and steam inlet velocity. The Lagrangian method is used, in which first the steam flow field is developed using the CFD code FrontFlow/Red and then the particle motion is simulated. It was found that the reduction of the plate spacing can improve the steam dryer performance without a significant increase of pressure drop, the wane pitch has a value after which the steam dryer performance is not significantly improved, the number of wanes of 1.5 was selected at this point, however, a more detailed model is necessary to arrive at the final conclusion. The optimum steam inlet velocity should be found using a detailed economical assessment. Velocities between 2 and 4 m/s seem to be reasonable to achieve good removal efficiency and keep the pressure drop at reasonable values.     

ENHS reactor power level enhancement possibilities

The ENHS thermal hydraulic optimization code was modified and applied to search for the maximum attainable power from a wide range of ENHS design options subjected to the following constraints: maximum permissible hot channel coolant outlet temperature of 600 °C, clad inner temperature of 650 °C and primary coolant temperature rise of either 150 °C or 90% of the theoretical limit for accelerated corrosion rate. The T–H optimization variables include the intermediate heat exchanger number of channels, channel width and elevation; diameter of the riser and diameter of a flow-splitting shroud in the riser. It was found possible to increase the attainable power from the nominal 125 MW_th up to 311 MW_th for the reference core, 400 MW_th for a reference-like core having equilibrium composition fuel and 372 MW_th for a flattened power core with 9 plutonium concentration zones. A power level exceeding 400 MW_th may be achieved by flattening the power distribution of the equilibrium core or using nitride fuel with enriched nitrogen rather than metallic fuel. With forced circulation it is possible to operate the flattened power core at up to 532 MW_th corresponding to 223 MW_e.     

Study on lead–bismuth droplets'' generation and their removal by an electrostatic precipitator

The applicability of the electrostatic precipitator for the removal of lead–bismuth droplets generated in the direct-contact boiling lead–bismuth cooled fast reactor is investigated. A small apparatus in which argon gas bubbles through the pool of lead–bismuth and an electrode mounted in the test section is used. The ESP operating voltage was 1000 V. It was found that the removal efficiency of the electrostatic precipitator increases with time up to 96.5%. It appears that the probability of droplet removal is almost independent of the droplet size. There is a small increase in this probability for larger droplets, which is caused likely by the fact that the larger droplets travel at lower velocities. Otherwise the effect of velocity on the removal efficiency is negligible. The electrostatic precipitator current was decreasing during the experiment, which is probably caused by the reduction of the number of droplets in the test section as the electrostatic precipitator was getting more efficient. The electrostatic precipitator current was on the order of 7 μA. The experiment demonstrated the applicability of the electrostatic precipitator for removal of lead–bismuth droplets.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

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.     

Natural circulation experiments in a non-uniform diameter lead bismuth loop and validation of LeBENC code

Experimental studies are carried out on natural circulation in a Lead Bismuth Eutectic (LBE) loop. The loop mainly consists of a heated section, air heat exchanger, valves, various tanks and argon gas control system. All the components and piping are made of SS316L. The dissolved oxygen in the LBE is monitored online by an Yttria Stabilised Zirconia (YSZ) oxygen sensor and controlled during the operation of the loop. In this paper the details of the loop and experimental studies carried out with heater power levels varying from 900 W to 5000 W are described. The temperature range of LBE during the experiments was 200 °C–500 °C. The maximum heat loss in the piping is kept less than 20% of the main heater power. Steady state experimental studies are carried out at different power levels and the LBE flow rate was found to be varying from 0.095 kg/s to 0.135 kg/s. The analysis and results of the performance of the heat exchanger with air and water as the secondary coolants are also discussed in the paper. Transient studies were carried out to simulate various events like heat sink loss, step power change and secondary side coolant flow rate change and reported in the paper. In the start up experiments, where the flow is started from stagnant condition of LBE, the time required for starting of natural circulation is found to be 600 s, 400 s and 240 s with power level of 1200 W, 2400 W and 3000 W respectively. The results are compared with available correlation and prediction of computer code LeBENC.     

Compatibility of surface-coated steels, refractory metals and ceramics to high temperature lead–bismuth eutectic

Compatibility of cladding material with lead–bismuth eutectic at temperature higher than 650 °C is one of the most crucial issues for feasibility of lead–bismuth-cooled fast reactors with cycle efficiency as high as 40%. In order to search for corrosion-resistant materials with lead–bismuth eutectic at temperature higher than 650 °C, surface-coated steels, some refractory metals and various ceramics were tested by means of stirred-type corrosion test. Lead–bismuth was heated up to 700 °C electrically in an alumina crucible, and oxygen concentration in the lead–bismuth was adequately controlled by injection of argon, steam and hydrogen gas mixture into the lead–bismuth. Specimens of aluminum–iron-alloy-surface-coated steels, refractory metals and ceramics including SiC/SiC composites were immersed in the stirred lead–bismuth for 1000 h. It was found that the surface-coated steels showed good compatibility with the lead–bismuth due to formation of a thin and stable protection layer on the surfaces. Tungsten and molybdenum exhibited high corrosion resistance. On the other hand, niobium is not a reliable material for the high temperature LBE. SiC and Ti₃SiC₂ also exhibited high corrosion resistance. On the other hand, the physical performance of the SiC/SiC composite must be improved especially by minimizing the porosity.     

CARR堆自然循环瓣阀设计及验证

自然循环瓣阀是中国原子能科学研究院中国先进研究堆(CARR)堆芯非能动余热排出系统的关键设备.按照功能及性能指标要求,瓣阀的设计采用阻尼臂和重锤两个可分离的转动部件的结构形式,以实现瓣阀在不同工况下的非能动开启或关闭.本文对瓣阀各组成部分的结构和作用进行描述,介绍了瓣阀的设计计算、材料选择以及性能试验.性能试验结果表明,该设计满足要求.     

脉冲堆热工水力设计程序研制开发

介绍了中国核动力研究设计院自主开发的脉冲堆热工水力设计程序系统。它包括脉冲堆自然循环分析程序（MC－FLOW）、堆芯热工水力分析程序（MC－THAS）和脉冲堆瞬态分析程序（MC－TRAN）。采用原型堆的数据对程序进行验证，其结果表明：脉冲堆热工水力设计程序系统满足热工水力设计的要求，能够可靠地用于西安脉冲堆的设计。     

一种低温堆核电站的动态数学模型

针对研发的采用一体化布置、全功率自然循环的低温核反应堆电站,建立了一个可用于大功率运行范围控制系统仿真的动态数学模型.模型采用了六组缓发中子动态方程(考虑了慢化剂温度和燃料温度反应性负反馈)、集中参数的堆芯传热模型以及自然循环流动模型,重点考虑了主回路自然循环对堆芯内冷却剂和燃料棒之间的传热系数、主换热器换热系数、主回路时间常数的影响.仿真结果表明,模型能够正确反映低温堆核电站的主要动态特性,可用于电站控制系统仿真.     

The prospect of mox fuel based PbBi cooled small nuclear power reactors

Safety performance of MOX fuel based PbBi cooled small fast power reactors has been analyzed and discussed. Though the thermal conductivity of MOX fuel is not large relative to that of nitride or metal fuel, but by proper combination of relatively small power density and relatively large natural circulation it can compensate fuel temperature decrease with coolant temperature increase smartly during unprotected loss of flow accident. Under such condition, accident analysis discussed in this paper show that under unprotected total loss of flow (ULOF) accident the reactor can survive inherently using combination of reactivity feedback. For unprotected rod run out transient over power (UTOP) accident the MOX reactor can overcome external reactivity by smaller power increase compared to that of nitride fueled reactors case. In this case doppler feedback plays much more important role compared to radial expansion component. So the MOX fueled small power reactors discussed here can survive both UTOP and ULOF accident with still enough temperature margin.     

Polonium evaporation and adhesion experiments for the development of polonium filter in lead–bismuth cooled reactors

Fundamental experiments were performed to determine the adhesion characteristics of polonium to different metals and to develop a filter for polonium evaporated from neutron-irradiated LBE. The results of the first experiments suggested that adhesion characteristics are almost the same for stainless steel and nickel metal. The results of the preliminary experiments for a polonium filter suggested that stainless steel mesh with thin wires could effectively collect polonium evaporated from neutron-irradiated LBE. In the experiments, stainless steel wire mesh was used, but from the results of adhesion experiment, it is expected that the same effect can be obtained with wire mesh made of other kinds of metal.     

Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models.This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 Nuclear Power Plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. Some of the proposed modifications of the system, which would significantly decrease risks associated with accidents that are beyond the original design basis, are discussed. The design of the newer 213 model differs in many aspects. It incorporates the simple and original application of passive natural processes to limit the large-break loss-of-coolant accident post accident pressure. Other features of this containment design, such as complicated geometry, dependence on several mechanical devices and interlocks, and insufficient experimental evidence, lead to doubts concerning the operation of this containment under accident conditions. For the newer 213 model, current work is devoted mainly to safety assessment and verification of the containment design. Some information concerning the on-going work is provided in the paper.     

反应堆功率保成本控制器设计

针对反应堆功率控制系统设计中存在的建模误差以及实际运行过程中参数的不确定性,提出了一种新的功率保成本鲁棒控制器设计方法:运用线性矩阵不等式先求解出最优状态反馈保成本控制器,构造卡尔曼滤波器解决实际问题中状态不可测的问题,采用回路恢复的方法使总系统逼近目标反馈环传递函数.系统仿真结果表明,所设计的控制器实现了对设定功率的精确跟踪,解决了在堆模型不确定性、运行过程中参数不确定性下控制系统可能出现的不稳定和控制精度不够的问题,系统不但具有良好的鲁棒性,而且调节性能较好,满足堆功率调节控制的要求.