The orifice throttle pipelines with large pressure drop suffer damages from two aspects: high frequency vibration caused by cavitation and low frequency vibration caused by high flow velocity. In order to solve the vibration phenomenon of typical throttling pipe with large pressure drop in the system of nuclear power plant, the key hydraulic characteristics such as pressure drop, flow velocity, streamline and eddy current were simulated and analyzed for single-stage orifice throttling pipe. The negative pressure area was found in the downstream of the orifice, where cavitation occurred. Eddy current was formed due to the large local velocity caused by the orifice plate jet. Then, the throttle performance of multi-stage concentric orifice plates was evaluated by means of blockage pressure drop method. The cavitation damage was relieved greatly. But it was not eliminated, especially in the last stage orifice. The expanding type five-stage orifice plate was designed according to pressure drop stage-decreasing principle, with which the possibility of cavitation was eliminated but large pressure drop resulted in large flow velocity at the downstream of the first orifice plate. Multi-stage eccentric orifice plate was designed with the consideration both in cavitation characteristics and velocity distribution, which could eliminate the harm of cavitation and reduce the low frequency vibration caused by large flow velocity to a maximum extent. As a result, multistage eccentric orifice plates could be recommended as an optimized design scheme for the vibration control of the orifice piping with large pressure drop.

In the active magnetic bearing (AMB) system, the eddy current is generated during the rotation of the rotor, which brings about the AMB loss and eddy-current magnetic field. The eddy-current magnetic field will reduce the electromagnetic force and generate the resistance torque of the AMB. Basing on the AMB of the Primary Helium Circulator (PHC) in HTR-PM, two-dimensional (2D) and three-dimensional (3D) AMB models are built and analyzed with finite element analysis (FEA) in maxwell. The 2D FEA shows that the eddy-current loss and the resistance torque increase as the rotor speed increases, and the eddy-current magnetic field will affect the air-gap magnetic field and reduce the electromagnetic force. The 3D FEA shows that dividing the rotor in insulate sheets can reduce the eddy-current loss. The loss and the sheets thickness have a linear relationship, which is different from the theoretical analysis.

The active magnetic bearing (AMB) is a new kind of high-performance bearing which suspends the rotor with controlled electromagnetic force. It was chosen to support the rotor of the helium blower in HTR-PM instead of conventional bearings. The power losses in the active magnetic bearings compose of three components: copper loss, iron loss and windage loss. In this paper, the iron loss, which composes of the eddy current loss and the hysteresis loss, is researched. The power loss of silicon steel lamination (35H300) was measured. Experimental data was taken over a range of 50Hz to 25,000Hz (sinusoidal current) for several magnetic field intensities. According to the experimental data, the eddy current loss and hysteresis loss increase with the frequency. And the hysteresis loss in the silicon steel lamination occupies the major part when the frequency of current is low, however the growth rate of eddy current is much faster than that of the hysteresis loss. And the FEM calculation of power loss in the magnetic bearing, which rotor and stator are made from silicon steel lamination (35H300), is also presented. The result shows the core loss of magnetic bearing also follow the separation theory. We can separate the core loss of magnetic bearing into two parts: hysteresis loss and eddy current loss. It will be very useful to calculate the power loss in the magnetic bearing.

Aiming to simulate the dynamical response of a non-ferromagnetic conductive structure in a strong magnetic field, a code of finite element method (FEM) was developed based on the reduced vector potential (Ar) method and a step by step time integration algorithm. The electromagneto-mechanical coupling effect was taken into consideration by adding ν × B term in the eddy current governing equation to calculate the additional electric field induced by the movement of the structure. The hexahedral isoparametric element was adopted in this code in order to simplify the correspondence between the simulation of electromagnetic force and the dynamical response, which enables the application of the code developed by authors to more complicated structures. To verify the validity of the new numerical code, the benchmark problem (TEAM-16) as a simplified model of Tokamak vacuum vessel structure was simulated. By numerical results contrasted between the current code and the ANSYS software, the code was proved to be more effective than typical commercial codes for structural analysis of a magneto-mechanical coupling problem. The simulation results proved that the new code can improve simulation accuracy especially in case of a large external magnetic field. In addition, the magnetic damping effect was also discussed in the paper.

After review of defects in nuclear reactor plants, this paper exhibits novel electromagnetic (EM) non-destructive testing and evaluation (NDT & E) for defect and material characterisation of nuclear fuel materials. Through comparison with traditional method for nuclear material inspection, e.g. X-Ray and phase array, EM NDT&E show advantages in non-contact inspection. Several recent development EM NDT & E techniques are reported and several case studies are investigated and discussed. Results show pulsed eddy current (PEC) can measure the stress and defect via the normalised ΔB Z features. The bigger stress results in the larger amplitude on normalised ΔB Z response. Magnetic Barkenhasuen noise (MBN) can measure the strain status via the RMS of Barkhausen noise amplitude. The RMS value decreases as the increase in strain. Eddy current pulsed thermography (ECPT) can be used for welding detection and characterisation through ‘highlighting’ the welding in thermal images. ECPT and MBN methods are also compared with traditional X-Ray and phase array methods for the inspection of welding samples. The comparison shows the competitive capability of ECPT in welding detection and evaluation. This paper demonstrates PEC, MBN and ECPT as good candidates for nuclear fuel materials.

The present study investigates modeling of thermal fatigue crack by the finite element method from the view point of eddy current testing. Three artificially produced thermal fatigue cracks introduced into Inconel 600 specimens are prepared. Eddy current signals are gathered by a differential type plus point probe with 25 kHz and 100 kHz. Subsequent destructive tests show the profile of these cracks. In the numerical simulation thermal fatigue crack is modeled as a region with constant width, uniform conductivity and real profile. Results of the study combining the previous study on modeling of thermal fatigue crack in type 304 stainless steel specimen reveal that thermal fatigue crack, unlike stress corrosion crack, generally should be modeled as an almost nonconductive region regardless of frequencies. Furthermore the resistance of thermal fatigue crack defined as width divided by conductivity is almost unchanged, even though the appropriate conductivity and width are changing with the frequency.

To detect degradation in steam generator (SG) tubes, periodic inspection using non-destructive examination techniques, such as an eddy current testing, is a common practice. Therefore, it is critical to evaluate and validate the reliability of the eddy current technique for ensuring the structural integrity of the SG tubes. The eddy current technique could be evaluated by comparing the data estimated by the eddy current with the destructive examination data of field cracks, which would be both costly and labor intensive. A viable alternative to pulled tube data is to manufacture crack specimens that closely represent actual field cracks in laboratory environments. A crack manufacturing method that can be conducted at room temperature and atmospheric pressure conditions is proposed. The method was applied to manufacture different types of stress corrosion cracking (SCC) specimens: axial outer-diameter (OD) SCC for straight tubes, circumferential ODSCC and primary water SCC (PWSCC) at hydraulic expansion transition regions, axial PWSCC at the apex and tangential regions of U-bend tubes. To help the growth of SCC into the tube, corrosive chemicals (sodium tetrathionate) and tensile stress were applied. Eddy current and destructive examination data for SCC specimens were compared with the available field crack data to determine whether those SCC specimens are representative. It was determined that the proposed method could manufacture the representative crack specimens.

Steam generator tubes are critical components of a pressurized water reactor, and have a number of important safety functions. Failure of steam generator tubes can contribute significantly to nuclear power plant station unavailability. It is critical to ensure that steam generator tubes perform their intended safety functions for the intended period of time consistent with the licensing basis, including applicable regulatory requirements. This can be achieved through inservice inspections of steam generator tubes that use nondestructive evaluation methods to detect and characterize unknown defects in tubes that could potentially leak before next inspection, so that such tubes could be plugged. In-service inspection data may be highly uncertain in nature due to detection uncertainties and measurement errors associated with nondestructive evaluation methods. These uncertainties and errors, if not properly accounted for, can result in defect size and density estimates not representative of actual tube degradations. This paper presents a Bayesian approach that combines prior knowledge of tube defects with uncertain nondestructive evaluation data considering the detection uncertainties and measurement errors to provide posterior distributions of true defect size and density. An example application of the proposed Bayesian approach is presented to estimate true defect size and density distributions in steam generator tubes based on in-service eddy current evaluation data. The proposed Bayesian probabilistic approach helps improve health management of steam generator tubes, thereby enhancing the overall safety and operability of pressurized water reactors.

Weld repair sites in nuclear systems are very often located in high radiation areas in difficult to reach sites. Additionally it is usually a pre-requisite to shut down the reactor while the repair is performed, ideally during a scheduled maintenance outage. To minimize the risk of an extended outage it is of paramount importance to ensure the repair time is optimized and the technique is reliable. Technical solutions to complex remote weld repairs are rarely available “off the shelf”. While some commercially available products can be adapted to suit an application, the constraints of space, distance and radiation usually require novel and unique designs involving a variety of engineering skills. Of equal importance to the weld repair is the requirement to perform pre and post repair nondestructive examinations (NDE). The application of ultrasonic, eddy current and liquid penetrate examinations requires a similar level of engineering. This paper outlines some of the welding and NDE techniques developed to address unique repairs to nuclear systems.

The 1.2MV, 70ns FWHM induction cell is developed for a 3MV Induction Voltage Adder (IVA) accelerator with three series connected cavities through a high voltage, vacuum insulating transmission line (VITL) driving rod-pinch diode (RPD) for radiography. The experimentally measured maximum relative permeability of IVA used amorphous material under pulse excitation is consistent with saturation wave model with flux density changing rate dB/dt greater than 10T/μs up to 32 T/μs. The remanence ratio B r /B s is experimentally measured for IVA pre-annealed amorphous cores. Upon experimental results, cores in each cell are designed and determined according to current transfer efficiency and volt-second integral necessary for each induction cavity, and peak field preventing VITL bore negative surface from electron emission. The field analysis is carried out for designing VITL vacuum stack and oil cavity that contains cores and azimuthal transmission line with operation field stress being about 50% critical breakdown field. The prototype cell tests validate electric field safety and magnetic core performance. With magnetization inductance and resistance of eddy current calculated for inductive cell, the IVA accelerator circuit model is set up, and simulation predictions approximately according with experimental results are presented.

The Poloidal Field (PF) coils, made of superconducting Cable in Conduit Conductor (CICC), are important for keeping plasma inside the vacuum vessel of ITER Tokamak device. To ensure the safety of the PF coil, Non-Destructive Testing (NDT) has to be performed for each conduit tube of CICC, as it is the boundary of liquid Helium and supporting structure of the electromagnetic load. However, the complicated cross section shape of the PF conduit tube causes difficulty in its NDT. Aiming at solving the NDT problem of the PF conduit tube, two approaches are proposed and their validity is demonstrated in this paper. One is a hybrid strategy of the ultrasonic phased array and the eddy current array technology. The other is of the immersed ultrasonic testing method. The experimental results reveal that both approaches are promising for the NDT of ITER conduit tubes though the later one is more efficient.

J-PARC (Japan Proton Accelerator Research Complex) is a multi-purpose research facility for materials and life sciences, nuclear and particle physics, and nuclear engineering with extremely high power proton beams of 1 MW. The accelerator complex consists of a 400-MeV linac, a 3-GeV Rapid Cycling Synchrotron (RCS), and a 50-GeV Main Ring synchrotron (MR). Its goals are to provide MW-class beams at 3 GeV and at several 10 GeV, while it is a challenge to localize and suppress beam loss to the level to allow hands-on maintenance of accelerator components. The RCS scheme is adopted to realize them, which is advantageous over conventional Accumulation Ring (AR) regarding less beam loss problems due to lower beam current and easier construction and operation of a linac. RCS, however, required various challenging technologies such as ceramic ducts to reduce eddy current effects, high field Radio Frequency (RF) system, and paint injection technique (an injection scheme to reduce phase space density of the beam) to reduce space charge effects. The linac has also unique technologies to minimize beam loss, such as compact electromagnet Drift Tube Quadrupoles (DTQ’s) to control beam envelopes precisely, and a fast beam suspending system in Machine Protection System (MPS) with Radio Frequency Quadrupole linac (RFQ). The beam commissioning of the linac started in Nov. 2006, and its design energy of 181 MeV in the first construction phase was achieved in Jan. 2007. RCS beam commissioning started in Sep. 2007 and the beam was accelerated to the designed energy of 3 GeV in Oct. 2007. MR beam commissioning started in May 2008, and the beam acceleration to 30 GeV was established in Dec. 2008. The first neutron and muon beams were produced in May and Sep. 2008, respectively, at Materials and Life science experimental Facility (MLF). The linac commissioning has resulted in very stable beam with short down time. RCS commissioning quickly achieved beam acceleration and extraction, and paint injections are being studied intensively. RCS recorded the highest beam power of 0.21 MW in Sep. 2008 with beam loss well localized at the collimators. The linac beam energy will be upgraded to 400 MeV with Annular Coupled Structure linac (ACS) in order to increase the beam power to 1 MW. In the second construction phase, upgrade of the linac with 600-MeV Super-Conducting Linac (SCL) for Accelerator-Driven nuclear waste transmutation System (ADS) and upgrade of MR energy from 30 to 50 GeV are planned.

Pitting corrosion is a serious form of degradation in steam generator (SG) tubing of some nuclear generating stations. The nature and extent of the pitting process is assessed through inspection programs, typically using eddy current probes. Although the data can be used in a probabilistic framework to predict the probability of a tube leak in future time intervals, the consideration of uncertainties associated with inspection tools is extremely important. The paper shows that the defect size data collected through inspections are significantly contaminated by probability of detection and measurement error such that the data present an incomplete picture of the defect population. The use of inspection data without accounting for inspection uncertainties in the analysis results in highly biased estimates of the leak probability. The paper presents a statistically correct likelihood method to account for inspection uncertainties and demonstrates its effectiveness through examples.

The basic method of the operational inspection of metal of heat exchanging tubes (HET) of steam generators (SG) is eddy-current multifrequency method all over the world. The greatest distribution was received variant with eddy-current testing (ECT) by use of a bobbing probe on the Russian nuclear power plants (NPP). Tubes with the defects which a subject to plugging are defined by results of lead operational ECT over the certain algorithm. SG resource is settled and replacement is required at achievement of a limit on number of the plugged tubes [1].

Steam Generator is one of the most critical components in nuclear power plants. It has of overriding importance from point of view of safe and reliable operation of the whole plant. Variety of degradation mechanisms affecting SG tube bundle may cause different types of material damage. In Paks NPP eddy current in-service inspection have been performed since 1988. In the year 1997 higher number of defected tubes were found in case of Unit#2, compared to results of the previous years. A medium term SG inspection program had been performed in the time period between 1998–2004. Based on the results of eddy current inspections high number of heat exchanger tubes had been plugged. Chemical cleanings of all steam generators were performed aiming to reduce the magnetite, copper deposits and corrosion agents acting on the surface of the tubes. Replacement of the main condensers had been performed to stop the uncontrolled water income caused by the relatively frequent leakages of the condenser tubes. Several tube samples had been cut from the first row of the tube bundles of different steam generators to study the effectiveness of the cleaning process and to determine the composition of deposits on the tube outside surface. Also several tubes with eddy current indications had been pulled out from the steam generators to determine the acting degradation mechanism. Examination of removed tubes can provide opportunity to check the reliability of eddy current inspection using bobbin coil. Also there were tubes pulled out form SG with existing cracks. From the year 2005 new inspection program had been started. As the first results of the new inspection program shows, there is only a few new indications had been found and there is no measurable crack propagation in case of existing indications. During the recent years feed-water collectors were replaced in case of all units of the power plant, because of material damage (erosion corrosion). The paper summarizes the results of eddy current in-service inspection of heat exchanger tubes, results of examinations of removed tubes and also deals with results of visual examination of the feed-water distributor system.

This paper discusses round-robin exercises using the NRC steam generator (SG) mock-up at Argonne National Laboratory to assess inspection reliability. The purpose of the round robins was to assess the current reliability of SG tubing inservice inspection, determine the probability of detection (POD) as function of flaw size or severity, and assess the capability for sizing of flaws. For the round robin and subsequent evaluation completed in 2001, eleven teams participated. Bobbin and rotating coil mock-up data collected by qualified industry personnel were evaluated. The mock-up contains hundreds of cracks and simulations of artifacts such as corrosion deposits and tube support plates that make detection and characterization of cracks more difficult in operating steam generators than in most laboratory situations. An expert Task Group from industry, Argonne National Laboratory, and the NRC have reviewed the signals from the laboratory-grown cracks used in the mock-up to ensure that they provide reasonable simulations of those obtained in the field. The mock-up contains 400 tube openings. Each tube contains nine 22.2-mm (7/8-in.) diameter, 30.5-cm (1-ft) long, Alloy 600 test sections. The flaws are located in the tube sheet near the roll transition zone (RTZ), in the tube support plate (TSP), and in the freespan. The flaws are primarily intergranular stress corrosion cracks (axial and circumferential, ID and OD) though intergranular attack (IGA) wear and fatigue cracks are also present, as well as cracks in dents. In addition to the simulated tube sheet and TSP the mock-up has simulated sludge and magnetite deposits. A multiparameter eddy current algorithm, validated for mock-up flaws, provided a detailed isometric plot for every flaw and was used to establish the reference state of defects in the mock-up. The detection results for the 11 teams were used to develop POD curves as a function of maximum depth, voltage and the parameter m p , for the various types of flaws. The POD curves were represented as linear logistic curves, and the curve parameters were determined by the method of Maximum Likelihood. The effect of both statistical uncertainties inherent in sampling from distributions and the uncertainties due to errors in the estimates of maximum depth and m p was investigated. The 95% one-sided confidence limits (OSL), which include errors in maximum depth estimates, are presented along with the POD curves. For the second round robin a reconfigured mock-up is being used to evaluate the effectiveness of eddy current array probes. The primary emphasis is on the X-Probe. Progress with the X-Probe round robin is discussed in this paper.

This paper illustrates the use of an “equivalent rectangular crack” approach to predict leak rates through laboratory generated stress corrosion cracks. A comparison between predicted and observed test data on rupture and leak rate from laboratory generated stress corrosion cracks are provided. Specimen flaws were sized by post-test fractography in addition to pre-test advanced eddy current technique. The test failure pressures and leak rates are shown to be closer to those predicted on the basis of fractography than on NDE. However, the predictions based on NDE results are encouraging, particularly because they have the potential to determine a more detailed geometry of ligamented cracks from which more accurate predictions of failure pressure and leak rate can be made in the future.

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned.