Analytical and experimental studies of tube/support interaction in multi-span heat exchanger tubes

Fretting damage of tubes in heat exchangers can be very costly and should be avoided. Information on vibratory responses and dynamic interaction between tubes and supports are prerequisites for understanding the relationship between frettingwear and tube vibration.A finite-element computation technique has been developed to predict the vibratory response and tube/support interaction of multi-span tubes.Experimental verficication of the computer prediction using a multi-span single tube apparatus has been performed in air and in water with clearance supports and with no-clearance supports. The natural frequencies, support impact forces and mid-span displacements of the experimental and analytical results were compared. The results are in fairly good agreement. An example is used to illustrate the application of the computation technique to a hypothetical heat exchanger tube. The life of the hypothetical heat exchanger tube is estimated based on the predicted support impact forces and existing wear data.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Shear-bending buckling analyses of fast breeder reactor main vessels

The purpose of this paper is to describe the computation results and the knowledge of the buckling analysis strategy. Since fast breeder reactor main vessels are thin shell structures, plastic shear-bending buckling is one of the most important problems. To clarify the buckling behaviour, we carried out many tests and numerical calculations. Based on the experience of those buckling analyses, available elements, mesh division, modelling of shape imperfections etc. are described. These results show that the numerical analysis can be a useful tool for evaluating buckling phenomena.     

中国实验快堆的安全特性

钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保...     

The BR-2 experimental fast neutron reactor

A brief survey is made of the design of the experimental fast neutron reactor and of its basic experimental and auxiliary equipment. The reactor was designed for physical experimentation with fast neutrons. The core is composed of plutonium rods; the lateral reflector is filled with depleted uranium. Heat is removed from the core by mercury and from the uranium reflector by air. The total rated power of the reactor is 150 kw of which about 100 kw is derived from the core.     

Modelling of fluidelastic vibrations of heat exchanger tubes with loose supports

The existence of gaps at tube supports necessitates time domain modelling of fluid forces to predict flow-induced vibrations and associated wear in heat exchangers and steam generators. This paper presents a new time-domain model for fluidelastic instability forces of tubes with loose-supports. In this model, the fluidelastic force, which is dependent on flow velocity and array geometry, is superimposed on the turbulence forcing function. The model was used to calculate the critical flow velocity, tube response, and tube/support interaction parameters, such as impact force and work rate. The critical velocity for linear cases was accurately predicted. The critical flow velocity for the loose support case was found to be sensitive to both the gap size and the turbulence level.     

An experimental study on natural circulation decay heat removal system for a loop type fast reactor

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.     

Influence of cyclic temperature transients on the creep behaviour of heat exchanger tubes

Due to thermal fluxes hot streaks exist in the coolant media of heat exchanger components. They cause alternating cyclically secondary stresses in the component walls which superpose on the primary stresses due to internal pressure or bending. Experimentally it was shown that hot streaks at high temperatures influence the creep behaviour very strongly. Dependent on the ratio of primary and secondary stresses the creep rate of the components is higher than the creep behaviour at the highest cycle temperature under primary stresses only.     

Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators. Part I. General corrosion state and morphology

The present work, constituting the first part of a series of two, deals with a systematic investigation of the general corrosion state of 22 heat exchanger tubes originating from different steam generators of the Paks NPP (Hungary). While the passivity of the inner surface of the stainless steel tube specimens was studied by voltammetry, the morphology and chemical composition of the oxide layer formed on the surfaces were analyzed by SEM–EDX method. Based on the measured corrosion characteristics (corrosion rate, thickness and chemical composition of the protective oxide layer) a strong dependence of these parameters on the decontamination history of the steam generators was revealed. It is well documented that the chemical decontamination carried out by a non-regenerative version of the AP-CITROX procedure does exert, on the long run, a detrimental effect on the corrosion resistance of steel surfaces. Therefore, process restrictions and modifications to minimize corrosion damages have be defined.     

Heat transfer in heat exchangers of sodium cooled fast reactor systems

The present paper describes the heat transfer in heat exchangers of sodium cooled fast reactors. Practical empirical correlations regarding heat transfer coefficients for intermediate heat exchangers (IHXs) and air coolers (ACs) were derived using test data obtained at the fast reactor ‘Monju’ and ‘Joyo’ and also at the 50 MW steam generator facility (50 MW SG). The correlation proposed by Seban and Shimazaki was applicable to estimate the heat transfer coefficients in both flows of IHX, i.e., primary and secondary flows, when the Péclet number was larger than 30. When the Péclet number for shell-side was small, the Nusselt number decreased as a function of the Péclet number. It was clarified that this characteristic was not caused by the heat conduction in flow direction. The heat conduction effect can be neglected even in the natural circulation conditions of the Monju plant. As for the heat transfer coefficient of AC provided in the secondary heat transport system of the fast breeder reactor, data in the above mentioned three facilities were evaluated. As a result, empirical correlations were derived for the average heat transfer coefficients of a large capacity finned air cooler made of stainless steel. These correlations could contribute to analyze the plant dynamics with better accuracy than before.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.     

The laboratory investigation of the experimental nuclear reactor VVR-S

Results are given in this paper of the laboratory investigation of the characteristics of the experimental reactor VVR-S undertaken with the aim of studying the neutron and physical parameters which are the most important for putting the reactor into operation and for its exploitation. As the result of the experimental work carried out, the critical mass and the maximum and operating fuel charges were found, the compensating capacity of the control and emergency rods was determined, the influence of the various factors (variation in the temperature of the water in the active zone, variations in the properties of the reflector, etc.) on the reactivity was studied, the distributions of neutron density with height and along the radius of the active zone were measured, and the operating time of the control rods was obtained.In conclusion the authors express their gratitude to T. N. Zubarev for discussion of the results and to O. I. Liubimtsev and I. V. Koptev for help in the work.     

Computational investigation of the effect of hydraulic and nondestructive test parameters on the strength reliability of the heat exchanger tubes of a steam generator

The results of calculations of the probability of a leak appearing in the tube band of steam generator in a VVéR-440 reactor system during operation are presented. The MAVR-1.1 computer code is used to calculate the probability of the formation of a leak and rupture of one of the heat exchanger tubes. The binomial distribution is used to determine the probability of the number of tubes that do not satisfy the plugging criterion. A leak in a tube bank is calculated as a sum of leaks in individual tubes. The probability of such a leak is calculated as a random sum. The calculations show that the parameters of test measures (pressure of the hydraulic tests, reliability of nondestructive testing for defects) and the sequence in which they are performed have a large effect on the failure probability of a tube bank during reactor operation. The computational results and the experience gained in operating steam generators show that the algorithm and the method developed for computing the leak probability could be helpful for estimating the strength reliability of heat exchanger tubes. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 216–221, April, 2007.     

Aspects of 238Pu production in the experimental fast reactor JOYO

Experimental determination of ²³⁸Pu in ²³⁷Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of ²³⁸Pu production from ²³⁷Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the ²³⁸Pu isotopic ratio. ²³⁸Pu production amount in the irradiated ²³⁷Np samples was determined by a radioanalytical technique. Aspects of ²³⁸Pu production were examined on the basis of the present radioanalysis. The ²³⁸Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu.     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators. Part II. Chemical composition and structure of tube surfaces

In the frame of a project dealing with the comprehensive study of the corrosion state of the steam generators of the Paks Nuclear Power Plant, Hungary, surface properties (chemical and phase compositions) of the heat exchanger tubes supplied by the power plant were studied by Mössbauer spectroscopy (CEMS), X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) methods. The work presented in this series provides evidence that chemical decontamination of the steam generators by the AP-CITROX technology does exert a detrimental effect on the chemical composition and structure of the protective oxide film grown-on the inner surfaces of heat exchanger piping. As an undesired consequence of the decontamination technology, a ‘hybrid’ structure of the amorphous and crystalline phases is formed in the outermost surface region (within a range of 11 μm). The constituents of this ‘hybrid’ structure exhibit great mobility into the primary coolant under normal operation of the VVER type reactor.