Correlation analysis for forced vibration test of the Haulien large scale seismic test program

The correlation analysis for forced vibration test of a 1/4 scale containment SSI test model constructed in Hualien, Taiwan, was carried out for the case of backfilled foundation. Prior to the correlation analysis, structural property was revised so that calculated fundamental frequency of fixed base condition was adjusted to that derived from test results. Correlation analysis was carried out by ‘Lattice Model’ which was able to estimate soil-structure interaction effects with embedment. The analysis results coincide well with test results and concluded that the mathematical soil-structure interaction model established by the correlation analysis is efficient to estimate a dynamic soil-structure interaction effect with embedment. This mathematical model will be applied as a basic model of simulation analysis for earthquake observation records.     

Preoperational vibration test of a primary heat transport system of the Canadian deuterium uranium reactor

During preoperational tests of Wolsong-2 nuclear power plant (NPP) which is a Canadian deuterium uranium (CANDU) reactor, vibration measurements were made on the primary heat transport (PHT) system. These measurements were evaluated by spectral methods to determine modal displacements and the modal stresses induced in the piping. The main aim of the measurement program is to confirm that the structural systems and components are adequately dimensioned for the operational vibration loads during the design life of the reactor. Structural analysis of the PHT system was performed using computer code to determine modal displacements and modal stresses analytically. The measurement results were compared with the analytically calculated resonance frequencies and modal values. A reasonable correspondence of the test and analytical results was achieved. The stresses evaluated on the PHT system were below the endurance limit for the material with a margin of safety.     

Load-displacement relation of reactor buildings subjected to earthquake forces

A major earthquake-resistant component of reactor buildings are solid earthquake-resistant walls of a box, cylindrical, frustrum or octagonal shape. In recent years, from the viewpoint of rationalization and economy of earthquake-resistant designs for reactor buildings, there have been movements in Japan which are trying to evaluate the safety of the buildings by confirming that the response value of them to a large earthquake force has a sufficient margin for the final stage. For this purpose it is necessary to model the load-displacement relation of each earthquake-resistant component of the building up to the final stage. This paper proposes calculation equations for the load-displacement relation based on the reduced scale model test results concerning the earthquake-resistant components of reactor buildings. The equations have been selected taking account of compatibility with the test results and practical use of the equations themselves.     

Construction materials for reactor buildings for decommissioning and recycling

A conceptual fluid–steel structure was studied to investigate the seismic characteristics of its use in the reactor building of nuclear power plants. The results of the earthquake response analysis of the conceptual fluid–steel structure showed that the structure had the same seismic safety ability as conventional reactor buildings. Applying the fluid–steel structure to a rector building, results in the following advantages: more elastic and light weight building materials, reducing the decommissioning wastes; the ability to recycle the structure materials because the fluid in the steel structure can be discharged the steel can be reused easily; the fluid in the steel structure has the possibility of reducing the seismic response of the structure by the sloshing damper effect. Further study is encouraged by this results.     

Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid–structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations.     

Forced vibration tests and simulation analyses of a nuclear reactor building

Forced vibration tests were carried out at the Hamaoka (BWR) Unit 4 R/B (reactor building) in Japan in April and May of 1992. Fundamental dynamic characteristics of the R/B were obtained, including its interaction with the adjacent T/B (turbine building) and the soil–structure interaction. Results for the preceding R/Bs are compared, and probable causes for fluctuations in the resonance curve around the 1st peak are discussed. Furthermore, simulation analyses of the fundamental dynamic characteristics of the soil–structure system were conducted, using a basic lumped-mass soil–structure model (lattice model), and strong correlation with the measured data was obtained. Other detailed simulation models were employed to investigate the effects of simultaneously induced vertical response and response of the adjacent turbine building on the lateral response of the reactor building.     

Compact ignition test reactor (CITR)

Design considerations have been developed for a compact ignition test reactor (CITR). The objectives of this tokamak device are to achieve ignition, to study the characteristics of plasmas that are self-heated by alpha particles, and to investigate burn control. To achieve a compact design, the toroidal field magnet consists of copper-stainless steel plates to accommodate relatively high stresses; it is inertially cooled by liquid nitrogen. No neutron shielding is provided between the plasma and the toroidal field magnet. The flat-top of the toroidal field magnet is 10 s. Strong auxiliary heating is employed. In one design option, adiabatic compression in major radius is employed to reduce the neutral beam energy required for adequate penetration; thiscompression boosted design option has a horizontally elongated vacuum chamber; illustrative parameters are a compressed plasma witha=0.50 m, R=1.35 m,B _T =9.1 T, and a neutral beam power of 15 MW of 160 keVD ⁰ beams. A design option has also been developed for alarge bore device, which utilizes a circular vacuum chamber. Thelarge bore design provides increased margin and flexibility; both direct heating with RF or neutral beam injection and compression boosted startup are possible. The large bore design also facilitates the investigation of high-Q driven operation. Illustrative plasma parameters for full use of the large bore area=0.85 m,R=1.90 m, andB _T =7.5 T.     

A computerized method for flow-induced random vibration analysis of nuclear reactor internals

The general analytical, numberical, and programming techniques of a computerized method for flow-induced random vibration analysis of nuclear reactor internal components is discussed. The statistical approach used is similar to that originally introduced by Powell and subsequently applied to predict the response of flat plates to homogeneous turbulent air flow. The input damping ratios and parameters related to the flow field are assumed to be known from experimental data, while the virtual mass and natural frequency shift effects due to hydraulic loading of the structure are included in the analysis. The latest numerical techniques developed for use with modern, high-speed digital computers are employed to evaluate the acceptance integrals, thus permitting the basic method to be applied to the vibration analysis of complex structures excited by inhomogeneous turbulent flow — a situation that is commonly encountered inside a nuclear reactor. The importance of computer program modulization and its relationship to overlays are discussed. Some representative predicted vibration amplitudes based on a typical pressurized water reactor design are given.     

Effect of small openings on strength and stiffness of shear walls in reactor buildings

A total of 26 wall specimens were tested to examine the effect of small openings on the strength and stiffness of shear walls in reactor buildings. The parameters tested were the shape, number and local arrangement of the openings, and the reinforcing method around the openings. Reversed cyclic loads were applied to the specimens, and their strengths and restoring force characteristics were compared with each other to understand the effects of these parameters. Based on the test results, two methods for predicting the shear strength of walls with numerous small openings are examined. One is a method to calculate directly the strength along assumed failure lines, and the other method allows estimation of the strength reduction factors for design, taking account of the effect of openings. Both methods are found to be useful to estimate roughly the strength of such shear walls. It is also shown that the stiffness of such walls could be evaluated using a combined multi-spring model. For the reinforcing methods around openings, a simplified reinforcing method is proposed and the effectiveness of the method is discussed by checking the contribution of the reinforcement to the wall strength.     

Investigation of the transition from forced to natural convection in the research reactor Munich II

The new research reactor Munich II (FRM-II), which is under construction at the Technical University of Munich (TUM), Germany, makes use of a compact reactor core consisting of a single fuel element, which is assembled of two concentric pipes. Between the fuel element's inner and outer pipe 113 involutely bent fuel plates are placed rotationally symmetric, forming 113 cooling channels of a constant width of 2.2 mm. At the Department for Nuclear and New Energy Systems of the Ruhr-University of Bochum, Germany, a test facility has been built in order to confirm the concept of the decay heat removal in the FRM-II, to acquire data of single and two phase natural convection flows and to detect the dry out in a narrow channel. All results have proved the possibility to cool the fuel element by means of natural convection.     

Comparative energetics of three fusion-fission symbiotic nuclear reactor systems

The energetics of three symbiotic fusion-fission nuclear reactor concepts are investigated. The fuel and power balances are considered for various values of systems parameters. The results from this analysis suggest that symbiotic fusion-fission systems are advantageous from the standpoint of economy and resource utilization.     

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.     

Numerical investigation of two control rod models for vibration noise in two dimensions

In two previous publications (Pázsit 1984, 1988) it was investigated how the spatial structure of neutron noise, induced by the vibration of localized absorbers, depends on (a) the strength of the rod (i.e. an investigation of the applicability of the weak absorber approximation), and (b) the way the vibrating absorbers are modelled. A quantitative analysis of both items was performed in 1-D. In 2-D, however, only analytical formulae were given.

The purpose of this paper is to present a quantitative analysis of the above two problems in 2-D. The results show that the domain of applicability of the weak absorber approximation is smaller in 2-D than in the 1-D case. On the other hand, the choice of the rod model plays a much less significant role in 2-D.     

Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead–bismuth

The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.     

Determination of a test section parameters for IRIS nuclear reactor pressurizer

An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered.     

Seismic response analysis of deeply embedded nuclear reactor buildings considering frequency-dependent soil impedance in time domain

In order to estimate the seismic behavior of deeply embedded nuclear power buildings, it is important to accurately transform the soil impedance in the frequency domain to the impulse response in the time domain. Although the transform is important for some nuclear buildings because they are deeply embedded in the soil, there are few practical and accurate methods at present. The author has proposed practical transform methods. In this paper, seismic response analyses considering frequency-dependent soil impedance in the time domain are shown. First, the formulation of the proposed transform methods is described. Then, the response analysis of a nuclear reactor building deeply embedded in inhomogeneous soil was performed considering the full matrix soil impedance as the example of practical problems. Through these analyses, the validity and efficiency of the methods were confirmed.     

Acoustic emission for on-line reactor monitoring: Results of intermediate vessel test monitoring and reactor hot functional testing

This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using that AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility.A vessel, designated ZB-1, has been tested under fatigue loading with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialprüfungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE.AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions.