System for high-level radioactive waste using microchannel chip — extraction behavior of metal ions from aqueous phase to organic phase in microchannel

In order to clarify the extraction behavior of U(VI) from aqueous phase to organic one in microchannel, we have carried out extraction experiments of U(VI) from HNO₃ aqueous solution of 3 M (M = mol/dm³) to 30% or 100% TBP phase in microchannel. From the results of extraction experiments, it was found that the extraction of U(VI) in microchannel could be performed in a short time for approximately 1 s with a good extractability in both organic phases of 30% and 100% TBP, and suggested that the other nuclides could be extracted with high extraction efficiency in microchannel. Furthermore, it is expected that the innovative and sophisticated nuclide separation systems can be developed by using microchannel extraction with selective extractants for specific nuclide.     

Use of poly(methyl methacrylate) in radioactive waste management: I. Radiation stability and degradation

In this study, in order to understand the possible use of PMMA in radioactive waste management as a solidifying agent, radiation stability of the PMMA was studied by gamma irradiations at two different dose rates of 1485 and 82.8 Gy/h. The total dose of irradiation was up to 523 kGy. Degradation nature was tested by studying the changes in mechanical and thermal properties with rate and total dose of irradiation. Ultimate tensile strength and toughness first increased and then decreased with total irradiation dose. Half value dose (HVD) for elongation was 148 kGy and it was 178 kGy for tensile strength at the dose rate of 1485 Gy/h. Half value dose was found from the extrapolation of experimental data as 228 kGy for elongation and 205 kGy for tensile strength at the dose rate of 82.8 Gy/h. The FTIR spectral analysis showed depolymerization degradation of polymer with irradiation. It was concluded from experimental results that PMMA can be used for embedding radioactive wastes.     

Flashing-induced density wave oscillations in a natural circulation BWR—mechanism of instability and stability map

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

可用于远程辐射探测系统的数据采集与监控平台的开发

针对目前企业、工厂的分布式布局以及远程监控的需要,同时基于产权的纠纷考虑,开发了充分利用现有可用的设备和技术,现场采用80x86硬件,软件系统MSDOS,应用软件自行开发,远程采用windows平台商用微机,利用oracle数据库读取配置,通过电话灵活监控远程现场的数据采集监控系统.该系统通过少许修改即可应用于多种场合,如机场监控、调度监控等.     

Forced convection heat transfer in rectangular ducts—general case of wall resistances and peripheral conduction for ventilation cooling of nuclear waste repositories

A numerical solution for laminar flow heat transfer between a flowing gas and its containing rectangular duct has been obtained for many different boundary conditions which may arise in nuclear waste repository ventilation corridors. The problem has been solved for the cases of insulation on no walls, one wall, two walls, and three walls with various finite resistances on the remaining walls. Simplifications are made to decouple the convective heat transfer problem from the far field conduction problem, but peripheral conduction is retained. Results have been obtained for several duct aspect ratios in the thermal entrance and in the fully developed regions, including the constant temperature cases. When one wall is insulated and the other three are at constant temperature, the maximum temperature occurs in the fluid rather than on the insulated wall. This maximum moves toward the insulated wall with increasing axial distance. Nusselt numbers for the same constant flux on all four walls with peripheral conduction lie in a narrow band bounded by zero and infinite peripheral conduction cases. A dimensionless wall conduction group of four can be considered infinite for the purpose of estimating fully developed Nusselt numbers to within an accuracy of 3%. A decrease in wall and bulk temperatures by finite wall conduction has been demonstrated for the case of a black body radiation boundary condition. Nusselt numbers for the case of constant temperature on the top and bottom walls and constant heat flux on the side walls exhibited unexpected behavior.     

Pyrochlore-structured titanate ceramics for immobilisation of actinides: Hot isostatic pressing (HIPing) and stainless steel/waste form interactions

A pyrochlore-structured titanate ceramic has been studied in respect of its overall feasibility for immobilisation of impure actinide-rich radioactive wastes through the hot isostatic pressing (HIPing) technique. The resultant waste form contains mainly pyrochlore (∼70%), rutile (∼14%) as well as perovskite (∼12%), hollandite (∼2%) and brannerite (∼1%). Optical spectroscopy confirms that uranium (used to simulate Pu) exists mainly in the stable pyrochlore-structured phase as tetravalent ions as designed. The stainless steel/waste form interactions under HIPing conditions (1280 °C/100 MPa/3 h) do not seem to change the actinide-bearing phases and therefore should have no detrimental effect on the waste form.     

Development of a simple reprocessing process using selective precipitant for uranylions — Engineering studies on systems for precipitating and separating

The simple reprocessing method based on precipitation using N-cyclohexyl-2- pyrrolidone (NCP) as a selective precipitant for U and Pu ions from dissolved solutions of spent FBR fuels has been developed. On the basis of fundamental studies on precipitation behaviors, we designed and built up the scaled-up laboratory equipments (1/20-scale of future plant capacity of 200 tHM/y) to evaluate technological and economical feasibility. This system, which mainly consists of a precipitator and a precipitate separator, should be operated continuously from economical reasons. From the experimental results, it was confirmed that the precipitator is capable of producing UO₂²⁺-NCP precipitate stably with stipulated residence time (approximately 30 min), and the precipitate separator has the highly efficient separation of precipitate from the slurry. Furthermore, the parametric experiments indicated that the rinsing operation increased the efficiency in decontamination of FP elements.     

对CANDU燃料组件结构的一维瞬态辐射程序开发与验证

在气冷CANDU式燃料组件之中,辐射换热也是不容忽视的一部分。特别是在出现了系统失压/失流事故时,辐射换热将会成为保证燃料安全的主要冷却手段。本文中针对CANDU式压力管编制了针对压力管几何条件下的一维辐射换热瞬态程序。程序中采用将燃料元件棒转化为同心圆环的方式简化辐射角的计算,并加入了隔层辐射模型,使模型更加贴近实际。采用分别将程序中的几个模块的计算结果与CFX计算结果对比的方式来达到程序验证的目的,验证结果显示程序RHTPB具有良好的表现,能够满足于反应堆安全计算的需要。     

辐射监测设备维修信息管理系统的设计与开发

介绍了一款辐射监测设备维修信息管理系统的开发过程及采用的方法.包括设计目标、设计构想和数据库的构建,软件的结构和主要界面.该系统减轻了操作人员的工作量,同时有效地提高了积累数据的利用效率.     

辐射加工级电子束吸收剂量量热计的研制

叙述了辐射加工级电子束吸收剂量量热计的设计原理,给出了用自行研制的石墨和盒式水量热计测定电子束吸收剂量的方法和结果。用不同吸收体材料和几何参数是热计测量参考材料的吸收剂量,所得结果在０．８％以内符合。并对现有的吸收剂量深度分布测量模体的关键结构作了改进。由测量结果刻度ＧＡＦ－ＤＭ－１２６０和ＦＷＴ－６０辐射显色薄膜得到的电子束吸收剂量响应曲线与＾６０Ｃｏγ射线响应一致。研制的量热计及其测试技术可作     

Ultimate heat sink for PHWRs—evaluation of existing methods and alternative proposals

Two methods are proposed for removal of decay heat after a reactor trip under loss of class IV power. One is based on natural circulation (thermosyphon) cooling while the other depends on the direct introduction of the standby cooling system (SCS) heat exchanger after the main pump coast-down. The present analysis shows that under bottled condition thermosyphon cooling is adequate to remove up to 10% full power without boiling and up to 12% power with boiling in the primary coolant channels. However, the direct introduction of the SCS obviates the uncertainties of thermosyphon and ensures positive flow driven by a pump available on class III power supply. This flow is so chosen that there is no boiling on the secondary side of SCS within the operable pressure range. The analysis shows that such an operation does not induce undue stresses in the equipment.     

从系统设计分析核电站辐射监测仪表的特点和发展趋势

基于核电站辐射监测系统设计,从辐射监测仪表的类型范围、功能、工程应用以及质量、价格和售后服务角度,对工作中接触较多的国内外主要辐射监测仪表供货商提供的连续辐射监测仪表的现状进行了比较.结合工程设计经验,分析了辐射监测仪表的特点和设计选型原则,探讨了辐射监测仪表的发展趋势和辐射监测仪表国产化应对策略,为新建核电工程辐射监测系统设计选用辐射监测仪表,实现辐射监测仪表制造国产化提供建议和参考.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Rules for design of nuclear graphite core components — Some considerations and approaches

The use of graphite as a structural element presents unusual problems both for the designer and stress analyst. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission.     

The new code “RCC-MR” — Rules for design and construction of LMFBR components

In 1978, Commissariat à l'Energie Atomique, Electricité de France, and Novatome decided to undertake a common effort to gather a complete collection of rules to apply for design of LMFBR components. The first issue of this work is now being published by AFCEN as the “RCCM” code. The preparation of the design rules used largely the experience gained in Superphenix components analysis, and the results of the large R&D program performed as a support for the design of this plant or at longer term perspective, coordinated by a scientific advisary council of AFCEN (Association Française pour les règles de Conception et de Construction des matériels des Chaudières Electronucléaires).     

The role of probabilistic methods in the development of regulations for nuclear power plant design — assessment and projections — in the Fed. Rep. Germany

Some views on present use and future potential of both reliability and risk analysis in reactor safety assessment and licensing are given. Although the deterministic approach is still dominating, the part of probabilistic methods in the process of regulating nuclear power plants is steadily increasing. Both methods are complementing one another.     

Use of γ Radiation from Short-Lived Fission Products for Detecting <Superscript>235</Superscript>U and <Superscript>239</Superscript>Pu

It is suggested that γ radiation with E _γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect ²³⁵U and ²³⁹Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 365–370, May 2005.     

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.     

Radiation transmission of heavyweight and normal-weight concretes containing colemanite for 6 MV and 18 MV X-rays using linear accelerator

Accurate measurements have been made to determine radiation transmission of concretes produced with barite, colemanite and normal aggregate by using beam transmission method for 6 and 18 MV X-rays with a linear accelerator (LINAC). Linear attenuation coefficients of thirteen heavy- and four normal-weight concretes were calculated. It was determined that linear attenuation coefficient (μ, cm⁻¹) decreased with colemanite concentration and increased with barite concentration in both types of the concretes.     

Probability-based load combinations for design of Category I structures — Overview of research program and recent results

This paper discusses the probability-based load combinations for the program dealing with the design of Category I structures, currently being worked on at Brookhaven National Laboratory (BNL) for the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission (NRC). The objective of this program is to develop a probabilistic approach for the safety evaluations of reactor containments and other seismic Category I structures subjected to multiple static and dynamic loadings. Furthermore, on the basis of the developed probabilistic approach, a load combination methodology for the design of seismic Category I structures will also be established.The major tasks of this program are: (1) establish probabilistic representations for various loads and structural resistance, (2) select appropriate structural analysis methods and identify limit states of structures, (3) develop a reliability analysis method applicable to nuclear structures, (4) apply the developed methodology to existing Category I structures in order to evaluate the reliability levels implied in the current design criteria, and (5) recommend load combination design criteria for Category I structures. When the program is completed, it will be possible to (1) provide a method that can evaluate the safety margins of existing containment and other Category I structures and (2) recommend probability-based load combinations and load factors for the design of Category I structures.At the present time, a reliability analysis method for seismic Category I concrete structures has been completed. By utilizing this method, it is possible to evaluate the safety of structures under various static and dynamic loads. In this paper, results of a reliability analysis of a realistic reinforced concrete containment structure under dead load, accidental pressure, and earthquake ground acceleration are presented to demonstrate the feasibility of the methodology.