福岛核事故后对我国核电厂应急控制中心抗震设防的思考

结合福岛核事故后对我国核电厂进行的核安全检查,分析了我国核安全法规关于核电厂应急控制中心的要求以及福岛核事故的经验教训,提出目前我国核电厂应急控制中心采用民用抗震设防标准进行抗震设防,无法保证在由地震引发的应急事故工况下应急控制中心的功能,应该适当提高其抗震设防级别。     

福岛核事故后核电厂安全改进行动分析

介绍了福岛核事故后世界上主要核电国家相继开展的核电厂安全检查、再评价行动,并得出相应的检查和测试结论。法国、美国和中国等国家分别提出了福岛核事故后改进核电厂安全的建议、要求和行动,并制定了具体工程措施：在极端外部事件的设防,严重事故预防和缓解,水、电、通风实体改进,限制严重事故下的放射性释放和应急准备等主要方面开展的安全改进行动,将会提高核电厂的安全水平并提升缓解严重事故的能力。反思福岛核事故,总结福岛核事故对核电安全技术改进的促进作用,对未来核电安全技术的发展进行了展望。     

利用福岛第一核电厂事故期间环境监测资料反推事故释放源项

2011年3月11日,日本福岛核事故导致放射性物质向大气环境的大规模释放。本工作利用大气数值预报模式WRFV2.2.1和大气弥散模式CALPUFF,对事故期间放射性物质的大气输运和弥散进行了模拟。应用大气释放源项的逆推算方法,结合单位释放率条件下的大气弥散模拟结果和环境监测数据,对福岛第一核电厂1到3号机组向大气环境释放的放射性核素总量进行了评估,推算的131I和137 Cs气载释放量分别为8.6×1016 Bq和8.6×1015 Bq。     

福岛核事故放射性物质大气释放及公众剂量评估进展

概述了日本经济产业省（METI）原子能安全保安院（NISA）、内阁府原子能安全委员会（NSC）、文部科学省（MEXT）、（独）日本原子能研究开发机构（JAEA）、东京电力株式会社（TEPCO）等发布的福岛第一核电站核事故放射性物质大气释放量、辐射环境监测及公众剂量评估的有关监测和估算结果,讨论了进一步评估此次事故的放射性后果需要关注的问题。     

Water chemistry technology – one of the key technologies for safe and reliable nuclear power plant operation

Water chemistry control is one of the key technologies to establish safe and reliable operation of nuclear power plants. Continuous and collaborative efforts of plant manufacturers and plant operator utilities have been focused on optimal water chemistry control, for which, a trio of requirements for water chemistry should be simultaneously satisfied: (1) better reliability of reactor structures and fuel rods; (2) lower occupational exposure and (3) fewer radwaste sources. Various groups in academia have carried out basic research to support the technical bases of water chemistry in plants. The Research Committee on Water Chemistry of the Atomic Energy Society of Japan (AESJ), which has now been reorganized as the Division of Water Chemistry (DWC) of AESJ, has played important roles to promote improvements in water chemistry control, to share knowledge about and experiences with water chemistry control among plant operators and manufacturers and to establish common technological bases for plant water chemistry and then to transfer them to the next generation of plant workers engaged in water chemistry. Furthermore, the DWC has tried and succeeded arranging R&D proposals for further improvement in water chemistry control through roadmap planning. In the paper, major achievements in plant technologies and in basic research studies of water chemistry in Japan are reviewed. The contributions of the DWC to the long-term safe management of the damaged reactors at the Fukushima Daiichi Nuclear Power Plant until their decommissioning are introduced.     

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – II: Feasibility study for the Fukushima Daiichi Nuclear Power Plant

The present article is the second part of a two-part article, examining the feasibility of the method proposed in the preceding part. In this method, the ratio of the core uncovered from water x _u, the ratio of the core flooded by water x _f, the ratio of the core slumped into the pedestal area of the drywell x _s, and the ratio of the injected water leaking before reaching the core x _wl are the four important uncertain parameters. The base case study shows that water injection via the core spray line is more effective to cool the uncovered core and to reduce the amount of cesium hydroxide (CsOH) released from the drywell. The sensitivity study is conducted by introducing the dimensionless decay heat N _qd, which combines the effects of x _f, x _s, and x _wl on the steam generation rate associated with the forced convection cooling in the reactor pressure vessel (RPV). The results show that the temperatures of the uncovered core and the other structures increase with N _qd. Consequently, the release rate of CsOH also increases with N _qd. The relationships of the measurable RPV wall temperature with the temperatures of the uncovered core and the structures as well as the release characteristics of CsOH are also examined.     

日本福岛核电站事故泄漏放射性核素漂移扩散状况分析

利用自主研发的放射性核素大气扩散模拟计算软件RADCES模拟计算了日本福岛第一核电站事故泄漏放射性核素的漂移扩散情况,给出了放射性核素云团的漂移扩散路径及到达CTBT国际监测系统部分核素监测台站的时间。结果表明,理论预测放射性核素云团到达时间与台站监测到异常样品的时间基本一致。     

Fukushima Daiichi fuel debris simulant materials for the development of cutting and collection technologies

Cutting fuel debris (solidified corium) is an important issue for the decommissioning of Fukushima Daiichi Power Station. The main reasons for developing and using suel debris simulants are presented. The relative merits of various types of materials (stainless steel, zircalloy, sintered ceramic, cast-fused zirconia, metal + ceramic, melted inactive simulants, prototypic fuel debris, irradiated fuel debris simulant) that can be used to test debris cutting have been assessed against criteria relevant for the cutting technique itself and also for (radioactive) aerosol and combustible gas generation. Simplified simulants can be used for the development of fuel debris cutting techniques but have limited representativeness so that melted inactive fuel debris simulant must be used to assess the cutting performance. Concerning combustible gas generation, zirconium plates provide an upper bound in term of underwater generation of hydrogen. Finally, for aerosol and dust generation, it appears that non-radioactive simulant cannot correctly represent the aerosol formed during cutting but prototypic fuel debris simulants, using depleted uranium are required. Laser cutting tests have been carried out with several types of simulant materials. Promising results were achieved in term of cutting ability both in air and underwater. Data have also been collected on the released aerosols. Nevertheless, confirmatory experiments with prototypic debris are still needed.     

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (II) – dynamic characteristics of KURION media for Cs removal in simulated contaminated water

The kinetic characteristics of the column were necessary property to be understood before actual operation. Hence, a functional small-scale zeolite column system was installed for conducting the experiments to understand decontamination behaviors. Each column has a 2 cm inner diameter and a 12 cm height, and 12 g of zeolite-type media was packed into the column. The column experiments were carried out with Kurion-zeolite, herschelite, at different feed rates of simulated water with different concentrations of Cs and sea salt. As expected from equilibrium ion-exchange isotherms obtained for KURION-herschelite, the adsorption of Cs is hampered by the existence of sea salt ratio. The difference in breakthrough behaviors can be ascribed to the difference in sea salt ratio. Above 1000 bed volumes, the adsorption rate of Cs was the same at a solution velocity of between 14 and 81 cm/min. Under the condition of a 3.4 wt% sea salt ratio, the performance of the media supplied by KURION was in the order surfactant modified zeolite < silver-impregnated engineered herschelite = herschelite (H). This result was suggested to evaluate the performance of KURION media on the actual columns.     

利用航测数据反推福岛核事故137Cs的释放量

日本福岛第一核电站发生事故后,造成大量的放射性物质释放。为准确评估事故的释放量,本文根据美国公布的航测¹³⁷Cs地面沉积浓度图和日本福岛第一核电站事故发生后观测的气象数据,利用拉格朗日烟团模式反推¹³⁷Cs的释放量,并通过计算估算日本福岛第一核电站核事故向大气释放的¹³¹I当量,约为1.07×10¹⁸Bq,估算结果与日本政府公布的估算结果接近。     

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – I: Basic concepts and model descriptions for the Fukushima Daiichi Nuclear Power Plant

This paper presents a simple approach for estimating the structure temperatures including the uncovered reactor core inside the reactor pressure vessel (RPV) and the release rates of fission products deposited in the RPV to the reactor building (R/B) at a certain time after the occurrence of a severe accident at a nuclear power plant (NPP). First, basic concepts are presented and then, a simplified steady-state heat balance model is proposed for estimating the temperatures of the uncovered reactor core and the upper structure in the RPV as well as the temperature of the RPV wall. In addition, models for estimating the revaporization rate of cesium hydroxide (CsOH) in the RPV and the leak rate of CsOH to the R/B via the drywell are also presented. The proposed approach is anticipated to be applicable to the damaged Units 1–3 of the Fukushima Daiichi NPP.     

竖直向下浸没式蒸汽直接接触冷凝流型研究

直接接触冷凝直接影响核反应堆安全壳抑压系统的性能。本文针对竖直向下浸没式直接接触冷凝流型开展实验研究,采用高速摄影仪记录不同湿阱过冷度条件下的蒸汽冷凝过程,依据流型特征划分了喘振、管外颈缩、向上球型脱落、向上T型脱落4种冷凝流型。研究了各流型对湿阱热分层的影响。实验结果表明,喘振流型和管外颈缩流型会增强湿阱内流体搅混,不易发生热分层,而向上球型脱落流型和向上T型脱落流型易引起热分层。理查森数（Ri）可作为流型转变的无量纲数,Ri<1时为管外颈缩流型,Ri>1时为向上脱落流型。