低功率加速管冷却系统的数值仿真设计

借助数值模拟的方法对低功率加速管的温度场进行模拟分析,为其设计出合理的冷却系统.分析数据显示该冷却系统的主要作用不是冷却而是使温度均匀化和降低环境影响,这可为类似装置冷却系统的设计提供参考.     

C波段6 MeV轴耦合驻波加速管的研制

为发展紧凑型电子直线加速器,本文研制了一支C波段轴耦合驻波加速管。该加速管包括3个聚束腔单元和9个均匀加速腔单元,总长度约284 mm。根据射频相位聚焦原理进行了初步物理设计,并对整管腔链进行等效电路分析及仿真优化,从而确定了尺寸参数,最后进行了冷测调配及高功率出束实验。基于该流程研发了C波段驻波加速管,其工作频率为5 713.6 MHz,束流能量可达6 MeV,脉冲流强为84.5 mA。     

中国实验快堆泵支承冷却系统温度场分析

中国实验快堆一回路泵支承套筒是承重设备,位于高温的热钠池中.为了限制套筒和套简内部冷钠腔室的钠温度,维持主泵正常工作温度,设置钠泵支承冷却通道.利用计算流体动力学技术(CFD),对泵支承冷却系统进行三维模拟,通过对泵支承冷却系统冷却流道和支承结构的数值传热分析,得到了该系统的温度场分布情况,验证了泵支承冷却系统的冷却能力.     

14MeV医用同源双模中能电子直线加速管的研制

医用同源双模中能电子直线加速管是影像引导放射治疗技术(Image Guide Radiation Therapy,IGRT)中的核心部件,为确保放射治疗直线加速器能够提供稳定和高品质的成像射束、双光子模式治疗射束以及多档电子射束,上海联影医疗科技有限公司研制了基于一种新型的能量开关技术的14 Me V医用双模驻波加速管。采用束流动力学程序Parmela对加速管整管的横向聚焦和纵向聚束进行了动力学设计分析,为优化加速管腔体几何结构提供了指标要求,最终利用电磁场仿真软件Superfish及CST(Computer Simulation Technology)优化腔体结构设计并得到了最优的微波参数。模拟计算结果表明,该加速管总长1.3 m,采用边耦合双周期?/2驻波结构,工作频率2.998 GHz,其输出束流能量可以实现多档可调,成像模式可输出低于3 MV的光子,治疗束可输出具有6 MV和10 MV两档的光子及4档能量电子束(最高能量可达14 Me V)。完成加工后,冷测结果与设计值符合得比较好,下一步将进行高功率微波老练。     

上海光源2998 MHz加速管波导耦合器的设计

上海同步辐射光源(SSRF, Shanghai Synchrotron Radiation Facility)直线加速器采用2998 MHz的加速管,这种加速管在结构形式上虽然与SLAC型2856 MHz加速管相似,但在尺寸上略有差异,需要重新计算.本文采用与实际调试加速管时相同的三频法、有限差分分析软件Superfish和HFSS对2998 MHz加速管耦合器进行物理设计,为实际工程设计提供指导.物理设计与实验测试结果非常吻合.     

非能动堆芯应急冷却系统试验

为了研究先进压水堆非能动堆芯应急冷却系统中各主要设备的行为和系统性能.中国核动力研究设计院在AC-600全压堆芯补水箱补水性能实验装置的基础上建成了非能动堆芯应急冷却系统试验装置在该试验装置上,根据不同的冷端破口直径、不同的压力平衡管和不同的自动卸压系统操作逻辑进行了一系列试验试验结果表明,不同的试验条件下,非能动堆芯应急冷却系统能够对堆芯进行冷却     

HSC加速器的初步大功率加速实验

混合单腔(HSC)型直线加速器是将一段4杆型射频四极(RFQ)加速结构和一段交叉指(IH)型漂移管(DT)加速结构混合到一个IH型腔里,利用IH型腔体具有的低功耗加速梯度高的优点,使得整个IH型混合腔的加速功率效率较其他加速结构腔体的更高。本文进行了低功率测试和初步大功率C6+加速实验。结果显示低功率测试结果与计算结果吻合较好,初步的大功率测试结果表明HSC加速器能满足下一步的高功率实验。     

中国实验快堆堆容器冷却系统全厂断电工况温度场分析

堆容器冷却系统是中国实验快堆(CEFR)-回路系统中的重要辅助系统之一,用于在各种工况下对反应堆堆容器进行冷却.本文利用国际通用的计算流体力学软件STAR-CD对CEFR堆容器冷却系统进行三维数值模拟,得到了在全厂断电事故发展过程中堆容器冷却系统的温度场和流场的瞬态分析结果,为相应部件的力学分析以及其它工况的分析提供了数据,对快堆优化设计和安全分析提供了重要的理论支持.     

低温永磁波荡器过冷液氮冷却系统的设计

低温永磁波荡器(Cryogenic Permanent Magnet Undulator,CPMU)是目前插入件技术发展的主要方向之一,其利用一些永磁材料,如钕铁硼(Nd Fe B)或镨铁棚(Pr Fe B)的磁场性能在低温下明显高于室温的特性来提高波荡器性能和光源束流品质,工作温区为50-150 K,需要冷却系统的冷却。CPMU冷却系统主要包括过冷液氮冷却系统和磁体阵列冷却回路。本文介绍了上海光源(Shanghai Synchrotron Radiation Facility,SSRF)CPMU过冷液氮冷却系统的设计方案和设计参数,进行了系统主要热负载的分析;对冷却系统中关键设备之一的过冷换热器进行了设计,并计算分析了过冷氮流经CPMU冷却系统的全程阻力损失,为系统另一关键设备液氮泵的选型提供依据。对CPMU过冷液氮冷却系统进行的在线测试表明,该设计满足CPMU样机的冷却需求。     

10MeV行波电子直线加速管的调谐与匹配

在电子直线加速管研制过程中,需对机械加工后的加速腔链及耦合器精细地进行微波调谐与匹配,以满足微波传输及加速电场的要求。本文介绍一支10MeV行波电子直线加速管的调配方法与流程,使用活塞探针法调谐均匀加速腔,采用谐振逼近法调谐非均匀加速腔,降低了调谐的复杂度。使用三频率法与移动负载法对耦合器进行了调配,调配结果满足指标要求。     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Basic concept of a near future BWR

This paper shows a basic concept of a near future boiling water reactor (BWR) aiming at evolutional safety and cost savings with minimum change from the current advanced BWR (ABWR). The plant output is uprated to 1500 MWe from 1356 MWe. This power uprate can bring about potential of 11% cost saving per MWe base. Safety improvement as a next generation large reactor is also achieved.

The advanced reinforced concrete containment vessel (ARCCV) is used for the containment vessel to improve safety for severe accidents. The peak pressure of the containment at severe accidents can be kept close to the design pressure. The advanced passive containment cooling system (APCS) is also provided and can accomplish no primary containment vessel (PCV) venting.

The advanced emergency core cooling system (AECCS) consists of four divisions in the front line. The advanced passive cooling system (APCS) is also provided. The combination of the four divisional emergency core cooling system (ECCS) and the passive safety system improves the plant performance in probabilistic safety assessment (PSA).

This plant concept is designed based on the heritage of the current ABWR. No more major research and development (R&D) are necessary. Therefore, construction and operation is possible in the early 2010s.     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core.     

小破口失水事故非能动系统瞬态特性研究

为了解先进压水堆小破口失水事故下非能动安全壳冷却系统、非能动堆芯冷却系统、非能动余热排出系统的瞬态响应特性,需开展小破口失水事故下反应堆冷却剂系统和安全壳的耦合响应特性研究。分析结果表明,小破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性与独立计算有较大差异,小破口失水事故下耦合分析得到的安全壳压力峰值小于独立计算。      

Specifics of RBMK core cooling in beyond design basis accidents

The most dangerous beyond design basis accidents for RBMK reactors, leading to the worst consequences, are related to the loss of long-term heat removal from the core. Due to a specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using control rods cooling circuit, depressurisation of reactor cooling system, supply of water into cooling system from low pressure water sources, etc. This paper presents the analysis of such heat removal by employing RELAP5, RELAP5-3D and RELAP/SCDAPSIM codes. The analysis was performed for Ignalina nuclear power plant with RBMK-1500 reactor. The analysis of result shows that the restoration of water supply into control rod channels enables to remove 10-30 MW of the generated heat from the reactor core. This amount of removed heat is comparable with reactor decay heat in long-term period and allows to slowdown the core heat-up process. However, the injection of water to reactor cooling system is considered as main strategy, which should be considered in RBMK-1500 accident management procedure.     

美国核电厂循环冷却水系统运行实践及借鉴

对美国不同类型核电厂所采用的循环冷却水系统进行了概括和总结,给出了直流、自然和机械通风冷却塔、冷却池及一次和二次循环混合使用等冷却方式在美国若干核电厂址的应用实践,并提出了针对我国核电厂冷却水系统的可借鉴之处。     

FFR一次辅助系统阻塞计氮气冷却系统实验研究

文章论证了我国第一座快堆ＦＦＲ一次辅助系统阻塞计冷却系统的设计方案；并根据ＦＦＲ一次辅助系统中阻塞计的设计参数、工作条件，实验给出和模拟不同阻塞计维持功率的情况下，氮气冷却系统的温度特性曲线；分析了此氮气冷却系统的热交换能力，实验证明该氮气冷却系统能够保证ＦＦＲ一次辅助系统阻塞计以程控模工作。     

用RELAP5程序分析非能动堆芯应急冷却系统实验结果

先进压水堆的一个显著特点是非能动系统的高可靠性，评价这些系统的运行特性以及系统分析程序（如RELAP5等）的计算能力是非常重要的，中国核动力研究设计院设计建造了原理性的非能动堆芯应急冷却系统实验装置，并完成了相关实验研究，取得一批有价值的数据，本文用RELAP5／MOD3．2程序对实验过程进行了模拟分析。通过计算结果与实验结果的比较，初步评价了RELAP5／MOD3．2程序的计算能力。     

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.