Experimental tests on the applicability of a radioactive tracer technique for measuring coolant mixing in nuclear reactor fuel subassemblies

A previously reported radioactive tracer method for measuring coolant mixing hot channel factor is developed here. Feasibility of the technique is demonstrated with a test section simulating part of a reactor sub-channel. The test section consists of a tube containing three hollow rods simulating the fuel rods, mounted in a small hydraulic loop. Tests were carried out by injecting a ³²P activated phosphate into the main flow at Reynolds number ranging from 20 000 to 70 000. The concentration of the tracer was measured by detecting the β-rays of ³²P with three windowless Geiger-Müller counters, placed inside the hollow rods, β-particles reach the counters through openings in various axial positions along the length of the rods, covered by a thin stainless steel foil. Measured values of the concentration are reported and corresponding values of coolant cross-flow are calculated.     

Nonlinear stability analysis of a reduced order model of nuclear reactors: A parametric study relevant to the advanced heavy water reactor

In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.     

Handling spent nuclear fuel and radioactive wastes on the shore servicing base in Gremikha

The situation which has developed at the shore base in Gremikha involving fuel assemblies in the removed cores of water-moderated water-cooled reactors (first-generation submarines), which are located on an open site in containers and receiving chambers, and involving solid and liquid radioactive wastes present at the base is examined. Data are presented on the number of fuel assemblies and their technical state and on the state and amount of solid and liquid radioactive wastes. Suggestions on what should be done with the fuel assemblies and wastes are discussed. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 61–65, July, 2006.     

基于遗传算法的压水堆核电一回路稳压器机理建模与仿真

针对压水堆核电站一回路稳压器实际运行特性,根据能量守恒、质量守恒和动量守恒方程,考虑了喷淋流量、电加热器功率及安全释放阀的影响,建立了一个两相动态非平衡的稳压器机理模型.为提高模型精度,采用遗传算法对该模型的参数进行优化,用以得到一组模型的最优参数.将参数优化算法应用于某900 MW核电站稳压器仿真实例,通过模块封装组建成稳压器压力仿真模型,并与核电厂提供的对应数据做了比较验证了建模方法的正确性及优化方法的有效性.     

Methodology to optimize radiation protection in radioactive waste disposal after closure of a disposal facility based on probabilistic approach

We propose a methodology to optimize radiation protection in radioactive waste disposal after the closure of a disposal facility based on a probabilistic approach. In this methodology, a set of alternative options of the disposal system design with the associated uncertainties estimated through a probabilistic approach are developed. Then, the methodology evaluates the advantages and disadvantages of each option for the optimization of radiation protection, in which it is possible to determine the compliance with a dose constraint of 0.3 mSv/y and to collect information for making a decision on the optimization. In particular, the obtained information on the exposure, which is the mode and width of the dose distribution, can be converted into information on engineering measures and site conditions for the options. This process allows us to discuss which option should be selected as the optimal option by considering the balance between the exposure and the engineering, economic, and social feasibility of the option. This methodology is helpful for providing clear reasons why an optimal disposal system design is selected by combining quantitative information on the exposure and the feasibility of each option.     

Reduction in radioactive discharges to the atmosphere and study of water deaeration practice in the primary loop of the VVR-M reactor

The function and operating principles of the closed deaeration loop of the reactor are described, with an account of operating experience of the deaeration loop during the first two years and the results of studies of deaeration management of the primary-loop water. The deaeration system in the VVR- M reactor is reported as not being the only means for removing the detonating mixture from the primary loop water and inhibiting the formation of gas bubbles containing detonating mixture in the loop piping.The closed loop for deaeration and combustion of detonating mixture designed and fabricated for the VVR-M reactor of the A. F. Ioffe Physics and Engineering Institute of the USSR Academy of Sciences has reduced to a mere tenth the amount of radioactive gases vented to the atmosphere.Translated from Atomnaya Énergiya, Vol. 19, No. 6. pp. 517–521, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

我国核电厂流出物监测和辐射环境监测标准体系研究

为了适应我国核电快速发展新形势下环境辐射安全监督管理要求,并更好地指导核电厂的流出物监测和辐射环境监测工作的开展,需要对现有监测标准体系进行梳理和分析,推动标准体系的完善。本文分析了目前我国核电厂流出物监测和辐射环境监测标准的技术要求,提出了我国现行标准体系需要补充和修订的标准清单及修订和完善的建议,为加强和完善我国核电厂流出物监测和辐射环境监测的监督管理提供参考借鉴。     

Information-analytical system for tracking the rehabilitation of a temporary storage facility for spent nuclear fuel and radioactive wastes in the village of Gremikha

An information-analytical system for solving various problems facing developers and implementing agencies at all stages of work has been developed at Russian Science Center Kurchatov Institute as part of a project for ecological rehabilitation of a temporary storage site located in the village of Gremikha for nuclear fuel and radioactive wastes. The information-analytical system accumulates all information on the project: technical characteristics of the objects, cartographic information, documentation, and information on the participants of the project, technologies, and equipment. The centralized storage of the initial data and the data obtained from examinations makes its use effective for all participants of the project. Analysis of the stored information and the formation of structured sets of data are necessary for planning further work on the project and choosing concrete design solutions, exchanging information between participants in the work, and providing information to state organs and the public.     

On the change in the color and transparency of glasses when bombarded by gamma rays from a Co60 source and in a nuclear reactor

Glasses of various types are widely used in radiation shielding. The problems involved in the effect of different types of radiation on glasses and the choice of glass with greatest radiation stability become important in this area. The present article deals with these questions.Results of experimental investigations where glasses of different types were subjected to bombardment from a Co⁶⁰ source and to bombardment in a nuclear reactor are discussed. Glasses of different composition, products of several different glassworks in the USSR, were tested. Techniques for enhancing the radiation stability of the glasses are recommended in this article. Those characteristics of glasses which provide a fairly complete picture of the radiation response of the materials, e. g., index of stability, index of saturation, minimum transparency, coefficient of rate of darkening, are introduced.In conclusion, I would like to express my acknowledgment to G. Ya. Vasil'ev and Yu, Ya. Zhelazova for their kind assistance in performing the experimental research discussed herein.     

Current perspectives on nuclear power plant risks and the use of risk-based information in regulation

For the past five years, the U.S. Nuclear Regulatory Commission has supported extensive studies of severe accidents. One outcome of this work is a set of advanced method for analyzing the probabilities, source terms, consequences, and risks of such accidents. These methods are being applied to a set of six U.S. commercial nuclear power plants, covering a wide spread of nuclear steam supply systems and containment designs. This work is to be documented in the Reactor Risk Reference Document, NUREG-1150, and supporting contractors reports. The methods being used for NUREG-1150, and some initial plant results, are briefly described in this paper.     

A decision supported system based on the combination of fuzzy expert estimates to assess the financial risks in high-level radioactive waste projects

This paper describes the technical aspects of the implementation of a Decision Aid System developed at ONDRAF/NIRAS/NERAS, the Belgian radioactive waste management agency. The proposed system is intended to deal with the economic and financial uncertainties inherent to Radioactive Waste Management (RWM) and especially to High-Level-Waste (HLW) repository projects. The economic calculus of such projects are particularly difficult due to many uncertainties that still exist regarding the final design, the eventual costs and the realization schedule. The general approach mainly relies on fuzzy logic rule-based systems to infer contingency or margin factors in such uncertain contexts. The parameters of the fuzzy models are identified based on preliminary rough estimates of the financial risks. This calibration enables to build a default system that can be further refined by the introduction of experts advices. The implemented software allows to introduce new information when available with the possibility to adjust the experts credibility.     

Calculation of the depletion of a radioactive plume in the atmosphere and subsequent exposure due to deposited activity

An efficient method has been developed for calculating the depletion, by dry deposition and washout to the ground, from a plume of fission products in the atmosphere, and for determining the isotopic composition of resulting gamma and beta activities in the ground deposit at any selected point downwind. By integration over the pattern of material on the ground, gamma and beta exposure rates may be derived as a function of the ageing time of the deposit. Elements in the plume are assigned to one of up to four ‘groups’; elements within each group being assigned a particular deposition velocity and washout coefficient.Calculations have been performed, for a notional MAGNOX depressurization accident, which illustrate the sensitivity of ground gamma spectra and exposure rate to the use of isotope-dependent deposition velocities. Circumstances under which it may be necessary to perform a full two-dimensional integration over the pattern of material on the ground are discussed.     

A second order turbulence model based on a Reynolds stress approach for two-phase boiling flow and application to fuel assembly analysis

High-thermal performance PWR (pressurized water reactor) spacer grids require both low-pressure loss and high critical heat flux (CHF) properties. Numerical investigations on the effect of angles and position of mixing vanes and to understand in more details the main physical phenomena (wall boiling, entrainment of bubbles in the wakes, recondensation) are required.In the field of fuel assembly analysis or design by means of CFD codes, the overwhelming majority of the studies are carried out using two-equation Eddy Viscosity Models (EVM), especially the standard K-? model, while the use of Reynolds Stress Transport Models (RSTM) remains exceptional.The simulation of swirling flow generated by the mixing vanes plays an important role for the prediction of the CHF for the fuel assemblies. For this reason, according to [14] and [Mimouni et al., 2009b], rotation effects and RSTM model are more specifically addressed in the paper.Before comparing performance of EVM and RSTM models on fuel assembly geometry, we performed calculations with simpler geometries, the DEBORA case and the ASU-annular channel case. ASU-annular channel case has already been addressed in [14] and [Mimouni et al., 2009b].Then, a geometry closer to actual fuel assemblies is considered. It consists of a rectangular test section in which a 2 × 2 rod bundle equipped with a simple spacer grid with mixing vanes is inserted. The influence of the turbulence model on target variables linked to CHF limitation will be discussed. Moreover, the sensitivity to the mesh refinement will be particularly examined. The study of this case is a further step towards the modelling of the two-phase boiling flow in real-life grids and rod bundles.     

In-service monitoring and fractographic inspection of ultrasonic indications on a feed water suction pipe of the Gundremmingen nuclear power plant

Radiographic and ultrasonic examinations of three pipes of the Gundremmingen nuclear power plant revealed cracks with a maximum depth of about 2 mm on the inside of the pipes with a wall thickness of 7.1 mm. For continued service until replacement two representative areas were monitored for crack growth by means of the potential drop method. As crack growth during the last year of service can be excluded by metallographic examinations after replacement the result of monitoring by means of potential probes – in relation to its resolution capacity no crack growth in the period under monitoring – is confirmed. A final evaluation based on the actual crack depths results in minimum resolution of 0.2 and 0.5 mm in crack depth. The suitability of the potential drop method for in-service monitoring of existing indications for crack growth – even until replacement of the component, if necessary – could thus be demonstrated convincingly.