增强IVR有效性的堆内注水策略研究

熔融物堆内滞留（IVR）是一项核电厂重要的严重事故管理措施,通过将熔融物滞留在压力容器内,以保证压力容器完整性,并防止某些可能危及安全壳完整性的堆外现象。对于高功率和熔池中金属量相对不足的反应堆,若下封头形成3层熔池结构,则其顶部薄金属层导致的聚焦效应可能对压力容器完整性带来更大的威胁。本文考虑通过破口倒灌及其他工程措施实现严重事故下熔池顶部水冷却,建立熔池传热模型,分析顶部注水的带热能力,建立事件树,分析顶部注水措施的成功概率及IVR的有效性。结果表明,通过压力容器内外同时水冷熔融物,能显著增强IVR措施的有效性。     

ACP600主蒸汽管道破裂事故的应对措施研究

针对ACP600取消高压安注系统和浓硼箱、使用一体化钆为可燃毒物、采用Mode-C运行与控制模式等设计改进导致主蒸汽管道破裂（MSLB）事故安全裕量降低的不利情况，对先进三代核电厂ACP600的MSLB事故进行分析研究。为避免MSLB事故下反应堆重返临界后堆芯功率峰值过高导致偏离泡核沸腾比（DNBR）低于限制值，分别从快速注入硼溶液和减缓堆芯冷却率的角度，评价不同的安注系统配置以及停运故障环路主泵对于缓解MSLB事故的作用。研究最佳的缓解方案，并提出增设“蒸汽管线压力低-3”信号停运故障环路主泵的设计优化建议。      

PRHRS隔离阀前后破口事故对非能动堆芯冷却系统的影响分析

对于AP型核电站小破口失水事故(SBLOCA)试验进程，国内外有较为一致的认识，但对于相同尺寸破口在不同破口位置对试验进程、非能动堆芯冷却系统的影响仍需进一步研究。本文利用大型非能动堆芯冷却整体试验台架ACME开展了非能动余热排出系统（PRHRS）隔离阀前后破口事故试验工况研究，并以堆芯补水箱（CMT）侧冷管底部破口事故工况作为对比工况。试验结果表明：ACME开展的PRHRS隔离阀前后破口事故模拟工况事故进程符合典型SBLOCA进程，堆芯始终处在良好的冷却状态，非能动堆芯冷却系统的安全性得到有效验证；相同破口尺寸工况下，不同破口位置对事故进程有一定的影响，其中破口位置对CMT液位、安注流量的影响较为关键。对比工况中，PRHRS设备换热量也有较大不同，冷管破口和隔离阀后破口工况较隔离阀前破口工况换热量更大，但PRHRS换热管内部流动换热机理需进一步研究。     

Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System

For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study.     

重力注水过程流动不稳定现象关键影响因素研究

重力驱动注水过程中由于流量较小以及蒸汽的积聚可能导致流动不稳定现象的发生，对核反应堆安全运行具有重要的影响。通过实验研究的方法，搭建了重力注水模拟实验装置，研究了不同蒸汽出口形阻、高位储水箱水位和加热棒初始温度下流动不稳定现象的变化规律。结果表明，重力驱动注水过程流动不稳定现象包含冷却水初次注入阶段、注入水逐出阶段和冷却水再注入阶段等。在一定冷却水初始温度、冷却水入口形阻以及系统压力下，蒸汽排出速度以及实验本体内筒顶部的聚集情况取决于蒸汽出口形阻，减小蒸汽出口形阻可加快蒸汽排放速度，压力峰峰值降低、振荡周期变长，有利于系统稳定；提高高位储水箱水位加快了冷却水注入速率，增加了加热棒被淹没率，降低了流动不稳定现象的发生次数和持续时间；随加热棒初始温度的升高，冷却水流量出现了波动向停滞的转变，流动不稳定现象发生的次数增加且持续时间加长。     

先进安注箱试验研究

先进安注箱与传统安注箱相比,可在安注的不同时期根据堆芯冷却需要自动转换安注流量,提高冷却液利用效率,同时可简化安注系统,节约建造成本。为获得所设计的先进安注箱水力学特性,在基于模化相似理论设计的先进安注箱试验回路上开展水力学试验研究,最终获得了箱体安注过程中安注流量、压力、液位、介质温度和水力学部件流阻系数等参数的变化规律。结果表明,本研究所针对的先进安注箱试验本体可实现安注流量的自动转换功能,且大小流量比在3.5∶1左右,安注总时间可达250s,与同类设备的安注性能指标相比具有一定的先进性。本试验结果不仅验证了先进安注箱结构设计的合理性,还可为反应堆安全分析提供输入性数据。     

Effect of edge magnetic island on carbon screening in the J-TEXT tokamak

The effect of externally applied resonant magnetic perturbation(RMP)on carbon impurity behavior is investigated in the J-TEXT tokamak.It is found that the m/n=3/1 islands have an impurity screening effect,which becomes obvious while the edge magnetic island is generated via RMP field penetration.The impurity screening effect shows a dependence on the RMP phase with the field penetration,which is strongest if the O point of the magnetic island is near the low-field-side(LFS)limiter plate.By combining a methane injection experimental study and STRAHL impurity transport analysis,we found that the variation of the impurity transport dominates the impurity screening effect.The impurity diffusion at the inner plasma region(r/a＜0.8)is enhanced with a significant increase in outward convection velocity at the edge region in the case of the magnetic island's O point near the LFS limiter plate.The impurity transport coefficient varies by a much lower level for the case with the magnetic island's X point near the LFS limiter plate.The interaction of the magnetic island and the LFS limiter plate is thought to contribute to the impurity transport variation with the dependence on the RMP phase.A possible reason is the interaction between the magnetic island and the LFS limiter.     

Study on modeling of spray cooling for spent fuel pool accidents

In Japan, spray equipment is prepared in spent fuel pools (SFP) in accordance with the regulatory requirements to mitigate fuel damage in the event that the water level of SFP cannot be maintained. In order to evaluate the spray coolability of fuel assemblies in SFP accidents, the spray cooling experiments were conducted under the SFP conditions. The experimental facility contains one mock-up BWR fuel assembly with full-length 7 × 7 heater rods in a mock-up SFP rack. The measured surface temperatures indicate that the spray injection results in the top-down quench and the precursory cooling, which are consistent with the spray-cooling mechanism that has been revealed by previous studies investigating reactor core spray. Further, the numerical simulations of the experiments were conducted using the TRACE code to examine the applicability of system codes for evaluating the spray coolability of SFPs. Although the TRACE calculation with a simple analytical model reproduced the top-down quench by spray injection as observed in the experiments, some qualitative differences were found between the experiments and calculations. The causes of these differences were revealed and the applicability of system codes were discussed.     

Thermal hydraulic behaviors during loss of shutdown cooling system at HTS opened under LLDS

When the CANDU6 nuclear power plant is operated at HTS opened under low level drained state (LLDS), loss of shutdown cooling (SDC) system is a potential accident that could challenge the core safety. Thermal hydraulic behaviors during loss of SDC system at HTS opened is studied with the CANDU6 nuclear power plant using RELAP5 code. Two basic cases, pump seal open and steam generator (SG) manway open, are analyzed. It is indicated that the core could keep safe for some time by intermittent flow established through the bubbles venting and fuel channel reflooding. The different vent size and location could result in different phenomenon. Two possible measures, the heat transport system (HTS) injection and reflux-condensation, are considered. The HTS injection could effectively remove the core decay heat for a longer time under conditions of coolant injection into HTS. For reflux-condensation, three cases are investigated. For 0 SG case, the HTS will be overpressure for a short time. For 1 SG case, the HTS will be overpressure some time later than 0 SG case, and analysis for 1 SG case with much more lower core decay heat shows that the HTS pressure will increase to a high level. The analysis of the case of 2 SG available for each loop shows that the decay heat can be removed effectively by reflux-condensation.     

Large Eddy Simulation of the Effects of Plasma Actuation Strength on Film Cooling Efficiency

In this article, numerical investigation of the effects of different plasma actuation strengths on the film cooling flow characteristics has been conducted using large eddy simulation(LES). For this numerical research, the plasma actuator is placed downstream of the trailing edge of the film cooling hole and a phenomenological model is employed to provide the electric field generated by it, resulting in the body forces. Our results show that as the plasma actuation strength grows larger, under the downward effect of the plasma actuation, the jet trajectory near the cooling hole stays closer to the wall and the recirculation region observably reduces in size. Meanwhile, the momentum injection effect of the plasma actuation also actively alters the distributions of the velocity components downstream of the cooling hole. Consequently, the influence of the plasma actuation strength on the Reynolds stress downstream of the cooling hole is remarkable. Furthermore, the plasma actuation weakens the strength of the kidney shaped vortex and prevents the jet from lifting off the wall. Therefore, with the increase of the strength of the plasma actuation, the coolant core stays closer to the wall and tends to split into two distinct regions. So the centerline film cooling efficiency is enhanced, and it is increased by 55% at most when the plasma actuation strength is 10.     

Analyses for Passive Safety of Fusion Reactor during Ex-Vessel Loss of Coolant Accident

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down a t about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time Period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident.     

Analysis of corium recovery concepts by the EUROCORE group

The objective of the EUROCORE (European Group for Analysis of Corium Recovery Concepts) Concerted Action is to obtain a clear view of the state-of-the-art for melt stabilisation as considered in accident management schemes and to better identify Research and Development (R&D) needs. Five different melt stabilisation concepts have been discussed: in-vessel retention with external cooling, core-concrete interaction with top cooling, ex-vessel spreading with top flooding, water injection by bottom flooding, and crucible concept with sacrificial material. For each concept, main unresolved problems are discussed in this paper and recommended R&D actions are outlined. The project started on 1 March 2000 and ended on 28 February 2002.     

A saw-tooth plasma actuator for film cooling efficiency enhancement of a shaped hole

This paper reports the large eddy simulations of the effects of a saw-tooth plasma actuator and the laidback fan-shaped hole on the film cooling flow characteristics,and the numerical results are compared with a corresponding standard configuration (cylindrical hole without the sawtooth plasma actuator).For this numerical research,the saw-tooth plasma actuator is installed just downstream of the cooling hole and a phenomenological plasma model is employed to provide the 3D plasma force vectors.The results show that thanks to the downward force and the momentum injection effect of the saw-tooth plasma actuator,the cold jet comes closer to the wall surface and extends further downstream.The saw-tooth plasma actuator also induces a new pair of vortex which weakens the strength of the counter-rotating vortex pair (CRVP) and entrains the coolant towards the wall,and thus the diffusion of the cold jet in the crossflow is suppressed.Furthermore,the laidback fan-shaped hole reduces the vertical jet velocity causing the disappearance of downstream spiral separation node vortices,this compensates for the deficiency of the saw-tooth plasma actuator.Both effects of the laidback fan-shaped hole and the saw-tooth plasma actuator effectively control the development of the CRVP whose size and strength are smaller than those of the anti-counter rotating vortex pair in the far field,thus the centerline and the spanwise-averaged film cooling efficiency are enhanced.The average film cooling efficiency is the biggest in the Fan-Dc =1 case,which is 80％ bigger than that in the Fan-Dc =0 case and 288％ bigger than that in the Cyl-Dc =0 case.     

Thickness effects of water overlayer on its explosive evaporation at heated metal surfaces

Molecular dynamics (MD) simulations have been employed to investigate the effect of the thickness of a water overlayer on the character of its ejection from a heated Au surface. The simulations are performed for five systems differing in the thickness of the water overlayer which was adsorbed on a metal substrate heated to 1000 K. For each system, an explosive evaporation occurs in the part of the water film adjacent to the metal surface and the upper part of the film is pushed off by the generated force. The average maximum temperature of the water film decreases as the film thickness increases. In contrast, the temperature achieved by the fast cooling due to the explosive evaporation shows an inverse trend. The significance of these model calculations to matrix-assisted laser desorption and ionization (MALDI) mass spectrometry is discussed.     

基于地球化学模拟分析杂质矿物对酸法地浸采铀的影响

为探明酸法地浸采铀过程中杂质矿物对铀浸出的影响,以分批浸出试验为基础,采用反应路径模拟探讨杂质矿物对铀浸出机制的影响,利用反应溶质运移模型探讨杂质矿物对铀浸出化学场时空特征的影响。模拟结果表明：方解石、黄铁矿、赤铁矿会与铀矿竞争酸,竞争由强至弱依次为方解石、赤铁矿、黄铁矿,其中黄铁矿在酸浸条件下溶解较弱,但生成的低价硫和亚铁离子能降低溶浸液的Eh值,导致铀浸出减少,赤铁矿在酸浸条件下因耗酸而降低溶浸液的酸度,但又促进黄铁矿的溶解,进而影响铀的浸出;在时间上,杂质矿物会使铀的溶解迁移存在不同程度的滞后,铀的溶解-沉淀旋回周期延长,整个模拟矿层沥青铀矿完全溶解时间更长,铀浸出速率降低;在空间上,杂质矿物会使模拟矿层中铀矿溶解范围减小,铀矿溶解-沉淀旋回过程中沉淀量增加,U（Ⅵ）迁移浸出所需时间延长,浸出铀的迁移累积峰值变化不同。     

DIVIMP Modeling of Impurity Transport in EAST

Simulations of carbon impurity transport in SOL/divertor plasmas with Ohmic heating on EAST tokamak were performed using the two-dimensional（2D）Monte Carlo impurity transport code DIVIMP.The background plasmas for DIVIMP simulations were externally taken from B2.5/Eirene calculation.Besides the basic output of DIVIMP,the 2D density distributions of the carbon impurity with different ionization states and neutral carbon atoms were obtained,the2D distributions of CII and CIII emissivities from C＋1and C＋2radiation respectively were also calculated.Comparison between the measured and calculated CIII emissivities showed favorable agreement,indicating that the impurity physics transport models,as implemented in the DIVIMP code,are suitable for the EAST tokamak plasma condition.     

Research on High Pressure Gas Injection As a Method of Fueling, Disruption Mitigation and Plasma Termination for Future Tokamak Reactors

High-pressure gas injection has proved to be an effective disruption mitigation tech- nique in DIII-D tokamak experiments. If the method can be applied in future tokamak reactors not only for disruption mitigation but also for plasma termination and fueling, it will have an attractive advantage over the pellet and liquid injection from the viewpoint of economy and engineering design. In order to investigate the feasibility of this option, a study has been carried out with relevant parameters for conveying tubes of different geometrical sizes and for different gases. These parameters include pressure drop, lagger time after the valve＇s opening, gas diffusion in an ultra-high vacuum condition, and particle number contour.     

反应堆压力容器直接安注热分布特性研究

我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。      

Overview of runaway current suppression and dissipation on J-TEXT tokamak

The main works on disruption mitigation including suppression and mitigation of runaway current on the J-TEXT tokamak are summarized in this paper. Two strategies for the mitigation of runaway electron (RE) beams are applied in experiments. The first strategy enables the REs to be completely suppressed by means of supersonic molecular beam injection and resonant magnetic perturbation which can enhance RE loss, magnetic energy transfer which can reduce the electric field, and secondary massive gas injection (MGI) which can increase the collisional damping. For the second strategy, the runaway current is allowed to form but should be dissipated or soft landed within tolerance. It is observed that the runaway current can be significantly dissipated by MGI, and the dissipation rate increases with the injected impurity particle number and eventually stabilizes at 28 MA s−1. The dissipation rate of the runaway current can be up to 3 MA s−1 by ohmic field. Shattered pellet injection has been chosen as the main disruption mitigation method, which has the capability of injecting material deeper into the plasma for higher density assimilation when compared to MGI. Moreover, simulation works show that the RE seeds in the plasma are strongly influenced under different phases and sizes of 2/1 mode locked islands during thermal quench. The robust runaway suppression and runaway current dissipation provide an important insight on the disruption mitigation for future large tokamaks.     

Simulations of the SOL plasma for FAST,a proposed ITER satellite tokamak

The first simulations with EDGE2D/EIRENE code of the SOL plasma in the FAST tokamak have been run for the basic H-mode scenario. Its similarity to ITER and relevance for DEMO bring interest to the study. Five different preliminary divertor designs have been examined by varying density at separatrix over the plausible range n_s,out = 0.7–1.0 × 10²⁰ m^?3. Margins exist for optimizing the design and minimizing the impurity injection rate even at the lowest density, with load below the safe limit of 18 MW/m² on the monoblock W targets, and to achieve a good degree of detachment at higher density. Both the plate tilting angle and the neutral dynamics are crucial factors. The detachment level can be significantly increased for the higher density scenario, while for the full non-inductive operation the injection of impurities will probably be necessary to reduce the heat load.