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期刊[J][序号]作者.题名[J].刊名,出版年,卷(期):起止页码.专著[M][序号]著者.书名[M].版本.译者,译.出版地:出版者,出版年:起止页码.专著中析出的文献[序号]析出文献的主要责任者.析出文献题名[M]∥专著主要责任者.专著题名.出版地:出版者,出版年:析出文献的起止页码.     

文后参考文献著录规则

期刊[J][序号]作者.题名[J].刊名,出版年,卷(期):起止页码.专著[M][序号]著者.书名[M].版本.译者,译.出版地:出版者,出版年:起止页码.专著中析出的文献[序号]析出文献的主要责任者.析出文献题名[M]∥专著主要责任者.专著题名.出版地:出版者,出版年:析出文献的起止页码.论文集[C][序号]作者.题名[C].出版地:出版者,出版年.论文集中析出的文献     

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期刊[J][序号]作者.题名[J].刊名,出版年,卷(期):起止页码.专著[M][序号]著者.书名[M].版本.译者,译.出版地:出版者,出版年:起止页码.专著中析出的文献[序号]析出文献的主要责任者.析出文献题名[M]∥专著主要责任者.专著题名.出版地:出版者,出版年:析出文献的起止页码.     

The design and processing of SONG library

Multi-group cross section library is the basic condition for lattice calculation, which provides the necessary nuclear data parameters. The multi-functional lattice code SONG(Si et al. in Nucl Power Eng 35:111–115,2014. doi:10.13832/j.jnpe.2014.S2.0112; Chen et al. in Nucl Power Eng 35:127–130, 2014. doi:10.13832/j.jnpe.2014.S2.0127) is developed to adapt to the research demands of different types of new reactors, especially the fourth generation(GEN-IV) reactors(Cerullo and Lomonaco in Nucl Fuel Cycle Sci Eng, 2012. doi:10.1533/9780857096388.3.333; Giorgio et al. in Energy Policy61:1503–1520, 2013. doi:10.1016/j.enpol.2013.06.101).The corresponding multi-group cross section library of SONG(SONGLIB) is well designed with much effort considering the next generation reactors' new feature in material, spectrum, burnup depth, etc. Therefore, the burnup chain(Maria and Jaakko in Nucl Sci Eng164:140–150, 2010. doi:10.13182/NSE09-14), energy group structure(Tholakele et al. in Ann Nucl Energy80:279–292, 2015. doi:10.1016/j.anucene.2015.01.038),and reaction path should be specially considered. The library SONGLIB is processed with the Evaluated Nuclear Data File(ENDF), nuclide data auxiliary processing code NJOYBAT(Si et al. 2014), and library management code MANLIB(Si et al. 2014). A series of verification work of SONGLIB is then carried out, and the calculated results are satisfied.     

美国《保健物理》（Health Physics）杂志英文摘要（2020年118卷第1期）

USE OF ROUTINE ENVIRONMENTAL MONITORING DATATO ESTABLISH A DOSE-BASED COMPLIANCE SYSTEM FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE Arthur S. Rood1,Helen A. Grogan2,H. Justin Mohler3,James R. Rocco4,Emily A. Caffrey5,Colby Mangini6,Jay Cartwright7,Travis Matthews7,Chris Shaw7,Mark E. Packard8,John E. Till9(1.K-Spar Inc.,Idaho Falls ID;2. Cascade Scientific Inc.,Bend OR;3. Bridger Scientific Inc.,Belgrade MT;4. Sage Risk Solutions,LLC,Aurora OH;     

文后参考文献著录规则北大核心CSCD

期刊[J][序号]作者.题名[J].刊名,出版年,卷(期):起止页码.专著[M][序号]著者.书名[M].版本.译者,译.出版地:出版者,出版年:起止页码.专著中析出的文献[序号]析出文献的主要责任者.析出文献题名[M]∥专著主要责任者.专著题名.出版地:出版者,出版年:析出文献的起止页码.     

2020年国外辐射防护相关会议

1.美国保健物理学会(HPS)第53届年中会议,2020/01/26-29,美国,马里兰州,贝塞斯达,https://hps.org/meetings/2.第10届国际同位素大会,2020/02/03-07,马来西亚,科伦坡,http://10ici.org 3.演讲“钋-210:一种有毒致癌物质”,2020/02/20,英国,索尔兹伯里,https://events.iop.org/polonium-210-poison-and-carcinogen-litvinenko-inquiry.     

美国《保健物理》(Health Physics)杂志英文摘要(2021年120卷第1期)

Determination of Organically Bound Tritium(OBT)Concentration in Fish by Thermal Oxidation and Liquid Scintillation Counting Method S.Rashmi Nayak 1,Renita S.D’Souza 1,Mohan M.Purushotham 1,Bharath Seraje 1,Dileep N.Blangat 2,Ravi P.Mana 1,3,Karunakara Naregundi 1(1.Centre for Advanced Research in Environmental Radioactivity(CARER),Mangalore University,Mangalagangothri-574199,India;2.Environmental Survey Laboratory,Kaiga Generating Station,Kaiga-581400,India;3.Formerly with Health Physics Division,Bhabha Atomic Research Centre,Trombay,Mumbai-400085,India).     

空间轨道单粒子翻转率预估的小样本法

Anewmethodwasproposedforpredictingsingleeventupset(SEU)rateofsemiconductordevicesinspaceorbitsbasedonBayestheory.TheSEUrateswerecalculatedforthesamekindofsemiconductordevicesbasedonsmallsample’sexperimentdata,byfittingthecurveoftherelationshipbetweentheupsetcross-sectionsofsemiconductordeviceandtheheavyion’sLET,andusingtheFOMmethod.ThepossibleintervalsoftheSEUratesinagivenconfidencewerepredicted,too.     

Yield of H2O2in Gas-Liquid Phase with Pulsed DBD简

Electric discharge in water can generate a large number of oxidants such as ozone, hydrogen peroxide and hydroxyl radicals. In this paper, a non-thermal plasma processing system was established by means of pulsed dielectric barrier discharge in gas-liquid phase. The electrodes of discharge reactor were staggered. The yield of H2O2 was enhanced after discharge. The effects of discharge time, discharge voltage, frequency, initial pH value, and feed gas were investigated. The concentration of hydrogen peroxide and ozone was measured after discharge. The experimental results were fully analyzed. The chemical reaction equations in water were given as much as possible. At last, the water containing Rhodamine B was tested in this system. The degradation rate came to 94.22% in 30 min.     

Experimental study on two-phase flow instability of natural circulation under rolling motion condition

Two-phase flow instability of natural circulation under a rolling motion condition is experimentally studied. The experimental results show the rolling motion induces a fluid flow fluctuation. At the trough point of the flow fluctuation, rolling motion can cause the early occurrence of natural circulation two-phase flow instability, and this case is defined as trough-type flow oscillation. The system stability decreases with increasing rolling amplitude and effect of rolling frequency is nonlinear. The complex overlap effect of trough-type flow oscillation and density wave oscillation can enhance the system coolant fluctuation; this case is defined as complex flow oscillation. Complex flow oscillation may be divided into two types: regular and irregular complex flow oscillations. Irregular complex flow oscillation is a transition type from trough-type flow oscillation to regular complex flow oscillation. Under the same thermal hydraulic conditions, the marginal stability boundary (MSB) of regular complex flow oscillation is similar to that of density wave oscillation without rolling motion, and the influences of rolling parameters on the MSB are slight.     

Integral boundary layer heat transfer modeling for Tehran Research Reactor

In the present work the validity of applying the Boussinesq approximation in the analysis of natural convection heat transfer along nuclear fuel plates with large coolant channel aspect ratios is evaluated. The Boussinesq approximation is introduced into the integral boundary layer equations governing the system to describe the velocity and temperature distributions of the coolant in the cooling channels. The fuel plate temperature is related to the adjacent coolant fluid temperature by a fundamental law in conduction heat transfer. Air and water are considered as fluids. The coolant flow is assumed to be fully developed which is a convenient assumption for coolant channels having large aspect ratios. Obtained results indicate that the Boussinesq approximation is merely applicable over a limited range of coolant channel outlet fluid temperatures. The use of this approximation produces conservative estimation of the critical plate power for air flow and non-conservative estimation of the critical plate power for water flow.     

The Fort St. Vrain high temperature gas-cooled reactor: X. Core temperature fluctuations

In October 1977, during the rise to power test program, the Fort St. Vrain high temperature gas-cooled reactor experienced the first of 37 fluctuation events involving primary coolant outlet temperature, nuclear detector signals, steam generator module gas inlet temperature and steam generator module main and reheat steam temperatures. In a 3 year investigation it was determined that the apparent cause of the fluctuations was movements of core components accompanied by periodic changes in bypass flows and crossflows of primary coolant helium. Installation of region constraint devices has eliminated fluctuations, but a single small primary coolant helium core outlet temperature redistribution is experienced routinely during rise to power.     

等速入口三喷口流动的温度振荡数值模拟

由于来自核反应堆堆芯不同通道冷却剂的温度存在差异,它们在上腔室混合后将产生复杂的温度振荡现象,可能干扰堆芯出口的温度测量,也可能造成周围固体结构的热疲劳破坏。本文基于FLUENT软件,采用大涡模拟湍流模型(LES)对三维平行三喷口流动进行数值模拟,并与实验结果进行了比较。结果表明：随着入口速度的增加,流体混合越来越剧烈,同时温度振荡的最大振幅逐渐减小,而相应的频率则逐渐增加。     

超临界压力水冷包层方案第一壁的热与应力分析

针对超临界水冷包层中第一壁的运行工况,利用数值计算软件ANSYS中CFX和Workbench两个模块对第一壁结构中的固体域和流体域进行数值分析研究。对比矩形管道和圆形管道内传热及热应力分布发现,矩形管道四个角域强化了壁面流体和主流流体的动量和热量的交换,使传热性能优于圆形管道,而四个角域的存在也造成了该处的应力集中,使结构材料的最大应力明显高于圆形管道。进一步研究冷却剂流向和冷却管道几何结构参数对第一壁结构温度场和应力场的影响发现,在ITER运行工况下,冷却剂流向影响很小,增大冷却管道直径和减小冷却管道最小壁厚均能改善第一壁结构材料中的最高温度,而这两个几何结构参数对第一壁应力的影响较为复杂。     

低压自然循环间歇泉流动不稳定性实验研究与RELAP5程序验证

对低压条件下自然循环回路内的两相间歇泉流动不稳定性进行了实验研究。同时,对RELAP5/MOD3.2程序计算低压自然循环间歇泉流动不稳定的可行性进行了验证分析。实验结果表明,低压条件下,间歇泉流动不稳定产生的根本原因在于有效驱动压头的周期性变化,与加热段内气液两相流动的形成-消失周期密切相关。下降段内流体温度越高,波动周期越短。实验数据与RELAP5/MOD3.2程序模拟计算结果符合较好,说明RELAP5/MOD3.2程序对模拟计算低压条件下自然循环间歇泉流动不定稳性具有较好的适用性。     

直流蒸汽发生器稳态与瞬态特性数值模拟

直流蒸汽发生器内二次侧流体被一次侧流体加热,其轴向热流密度分布与两侧流体运行参数密切相关。采用RELAP5/MOD4.0程序对直流蒸汽发生器的稳态与瞬态运行特性进行数值模拟研究,获取了直流蒸汽发生器热工参数的振荡特征,并进一步探究了一个周期内不同时刻的二次侧流体的质量流量、热流密度及管壁温度的瞬时轴向分布以及二次侧入口节流对流动振荡的影响。结果表明:对流加热工况下轴向热流分布主要取决于一次侧流体和二次侧流体之间的温差。当直流蒸汽发生器内发生密度波振荡时,一次侧出口温度及总加热功率均剧烈振荡。干涸点的位置大幅移动,且其附近的壁温强烈波动,可通过增大二次侧入口节流抑制流动振荡。      

Experimental investigation on heat transfer characteristics in a safety injection nozzle component in PWR

Thermal fatigue is a potentially significant degradation mechanism in Nuclear Power Plants (NPP). For the fatigue analysis, the thermal load information about components must be determined firstly. In this paper, an experimental study was carried out to obtain local fluid temperatures and local heat transfer coefficients for the safety injection nozzle component in reactor coolant system (RCS). In this mixing tee component a hot jet issues into a cold cross-flow stream from an oblique pipe and the turbulent mixing of two fluids induces local cycling stresses on the adjacent piping wall. Experiments were performed using a special-made heat fluxmeter, which can measure the mixed fluid temperature close to the wall and the heat transfer coefficient between the fluid and the wall. Plexiglass and metallic 1/9-scale mockups were manufactured for flow visualization and heat transfer tests, respectively. All tests were conducted at range of 0–40 for the jet-to-cross-flow velocity ratio. The flow visualization test has obtained general pattern of the flow and identified sensitive zones in the component where the jet and cross-flow interact intensively to cause thermal fatigue more possibly. In the heat transfer test, heat fluxmeters were positioned in the wall at these sensitive zones. The measurement results of temperatures and heat transfer coefficients have been discussed in detail in the paper. These experimental results allow us improving the state of knowledge of the thermal load to be used in the industrial mixing tees in operating for long lifetime assessment and for the design in the basic Nuclear Power Plants.     

Some Relations of Neutronic Noise with Fluctuation of Inlet Coolant Temperature and Vibration of a Control Rod Obtained by Simultaneous Measurements at KUR

Experiments on reactor noise were conducted at KUR. Depending on the operating condition of the reactor, the cause of the noise are classified into the following four types.

1. Zero-power noise source due to the branching process of fission neutrons and/or due to random bombardment of neutrons to the detector—under natural circulation of coolant and at essentially zero-power level.

2. Coolant temperature fluctuation due to natural convection—under natural circulation and at relatively high power level.

3. Flow induced vibration of shim control rods—under forced circulation of coolant and at low power level.

4. Fluctuation of inlet coolant temperature—under forced circulation and near the maximum power level.

Vibration of a spare shim control rod and fluctuation of inlet coolant temperature were measured simultaneously with the neutronic noise. Then the noise sources of the types (3) and (4) were verified. The vibration of a control rod has a broad spectrum in low frequency region besides the large peak at 14 Hz. The fluctuation of inlet coolant temperature is non-white noise and consists of large low frequency component. The theoretically predicted sink structures in the neutronic PSD relating to the transit time of inlet coolant temperature fluctuation through the core were not observed in the experimental results.     

直流蒸汽发生器稳态与瞬态特性数值模拟

In once through steam generators, the secondary fluid is heated by the primary fluid. Therefore, the axial heat flux distribution is related to the operational parameters of the fluids in both sides. The RELAP5/MOD4.0 code is adopted to perform the static and transient simulations of a once through steam generator. The temperature, thermal equilibrium quality and heat flux distributions under both heating modes are analyzed and compared. The oscillation behavior of the once through steam generator is discussed and the instantaneous axial profiles of secondary mass flow rate, heat flux as well as tube wall temperature corresponding to four representative times in an oscillation period are further investigated. Finally, the effect of increasing the secondary inlet throttling is studied. The results show that the shape of heat flux distribution under convective heating mainly depends on the temperature difference between primary fluid and secondary fluid. When density wave oscillations occur in the once through steam generators, the primary outlet temperature and total heating power oscillate intensely. The location of the dryout point moves greatly and the tube wall temperature fluctuates most violently in its vicinity. Flow oscillation could be suppressed by increasing the inlet throttling of the secondary side.