Strain-rate dependence of low cycle fatigue behavior in a simulated BWR environment

The strain-rate dependence of low cycle fatigue behavior of ASTM A533B low-alloy steel was investigated in a simulated BWR environment. Fatigue resistance of the steel was found to be closely dependent on cyclic strain rate in high-temperature water. A tortuous cracking morphology was dominant at high strain rate and typical fan-like or quasi-cleavage cracking patterns were frequently observed on the corresponding fracture surface. An entirely straight cracking morphology, however, became dominant at low strain rate. Evidence of crack arrest was found on the fracture surface rather than fan-like or quasi-cleavage cracking patterns. The above cracking behavior in simulated BWR water may be attributed to a strain-rate-induced change in the dominant environmentally assisted cracking (EAC) mechanism from hydrogen-induced cracking to film-rupture/slip-dissolution-controlled cracking.     

Figure of merit for the quality of ZrO2 coatings on stainless steel and nickel-based alloy surfaces

A figure of merit (FOM) has been developed to define the quality of ceramic (e.g., ZrO₂) coatings on metal and alloy [Type 304SS and Alloy 600] surfaces. Zirconia (ZrO₂) coatings were developed as a means of protecting the metal/alloy surfaces from stress corrosion cracking (SCC) in boiling water reactor (BWR) primary heat transport circuits, by inhibiting the cathodic reaction (reduction of oxygen and hydrogen peroxide) on the surface external to the crack. The distribution of pores in the coating plays an important role in corrosion prevention, such that the lower the porosity of the coating, the better the protection afforded to the system against SCC. Since the reactors operate at high temperature (e.g., at 288 °C under full power conditions), the temperature dependence of the FOM was investigated. The figure of merit (FOM) was developed by measuring impedance data over a wide range of frequency (10 mHz-5 kHz) at temperatures of 25, 100, 200, and 288 °C in hydrogenated, buffered solutions, with the hydrogen electrode reaction (HER) being used as a “fast” redox couple. An “equivalent circuit” analog was first developed from the bare surface impedance data and this analog was then employed in a second step to model the pore bottom in defining the pore distribution on the coated surface. A lognormal distribution (LND) of the pores was assumed and the parameters of the LND were determined using a constrained optimization technique to fit the model to the experimental data for the coated surface at different temperatures. The results suggest that, as temperature increases, the coating becomes more porous, making the substrate more susceptible to corrosion cracking. At 288 °C, 87% of the SS and 85% of the Ni-alloy surfaces become porous with the pore radius varying from 0.0001 cm to 0.01 cm.     

Application of advanced core process monitoring procedures in German power reactors

The nuclear reactor core design and the nuclear fuel management have been changed remarkable during the last few years. This development was initiated by increasing costs for the fuel recycling and nuclear waste storage. The fuel material, the fuel pellet fabrication, the fuel assembly structure and the core composition have been varied to get an effective fuel exploitation. Based on advanced core process conditions the reactor power and the fuel burn-up have been increased at German plants in recent years. Improved dynamic process monitoring procedures are required to get more information about the varied core process behaviour during the reactor operation. Since several years ISTec has been performed investigations to the process monitoring based on process signal measurements in German nuclear power plants. Using the standard instrumentation of the plants process signals have been measured and analysed by means of the digital data acquisition system SIGMA. The measured time signals are influenced by core process transients, global and local process fluctuations and by signal line transfer functions. Advanced time series analysis methods have been applied to separate different process effects in the multiple signal matrix. The separation of different process influences can improve significantly the information about the process condition in the reactor core.     

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades in boiling water reactors (BWRs) is important for safety assessment and for achieving flexible operation during the cycle. Characteristics that are affected significantly include distributions of the total fission (F_tot) and ²³⁸Ucapture (C₈) rates for controlled regions. Representative experimental investigations have been performed in the framework of the LWR-PROTEUS programme. In particular, the LWRPROTEUS I-2A experiments concerned the neutronics characterisation of a SVEA-96+ BWR assembly controlled with a hafnium (Hf) blade under full-density water moderation conditions. The current paper presents the comparisons of the measured F_tot and C₈ pin-wise distributions with a variety of stochastic and deterministic calculations: (a) MCNPX2.5 using recent nuclear data libraries (JEFF-3.1, ENDF/BVI. 8, and JENDL-3.3), (b) PHOENIX4 using ENDF/B-VI.3, (c) BOXER using JEF-1, (d) CASMO₄ using JEF-2.2, and (e) HELIOS1.6 using ENDF/B-VI.1. The calculation/experiment comparisons show standard deviations from 1.2% (MCNPX2.5) up to 1.9% (BOXER) for the prediction of the F_tot distribution, the highest individual discrepancy (7.6% with BOXER) being seen close to the “Hf-vertex.” The C₈ comparisons show systematically better agreement than those of F_tot, the lowest standard deviations being 1.0% (BOXER) and the highest only 1.4% (HELIOS). In addition, sensitivity studies highlight the greater importance of modelling aspects, compared with that of nuclear data libraries, for the achievement of satisfactory and validated F_tot and C₈ predictions.     

Work Hardening,Sensitization, and Potential Effects on the Susceptibility to Crack Initiation of 316L Stainless Steel in BWR Environment

After CBB testing, the structure and composition of surface oxide films on austenitic stainless steels in high-temperature water were examined by changing the metallurgical conditions in alloys and the concentration of dissolved oxygen and dissolved hydrogen in high-temperature water. SEM, X-ray diffraction, and Auger electron spectroscopy revealed that the water chemistry played a major role in the formation of the oxide film and in the crack initiation. The oxide film is composed of an outer layer formed with hematite and magnetite and an inner layer corresponding to the transition oxide/substrate, which is Cr-rich and Ni-rich. Dissolved hydrogen conditions reduced the thickness of the oxide and increased its nickel content. Dissolved oxygen conditions promoted the formation of numerous and bigger cracks compared with dissolved hydrogen conditions. Environmental conditions are preponderant compared with metallurgical conditions.     

Structural Material Anomaly Detection System Using Water Chemistry Data, (III)

A method to determine the electric conductivity of water continuously and directly at elevated temperature up to 300°C was developed which can be applied as a sensor for corrosion behavior foreknowledge and diagnosis systems using water chemistry data of BWR primary coolant. Complex impedance was measured between a couple of parallel platinum electrodes installed with a constant distance and dipped in the water. By analyzing frequency dependence of the impedance, the resistivity of the water between the platinum electrodes was estimated separately from the impedance caused by surface reactions on the platinum electrodes, which was the source of error in the measurement of electrical conductivity at elevated temperature.

Increase of necessary frequency to obtain the surface impedance at elevated temperature was evaded by the extrapolation of the frequence dependence of the impedance with calculations by using the data up to 100 kHz in which electric conductivity shows negligibly small dependence on the frequency of applied voltage.

The measured conductivity of pure water up to 300°C showed a good agreement with the calculation based on the dissociation data of water, which shows the applicability for the in-line monitor of electrical conductivity at elevated temperature.     

Thresholds for failure of high-burnup LWR fuels by Pellet Cladding mechanical interaction under reactivity-initiated accident conditions

ABSTRACT

To contribute to the future updating on the Japanese safety criteria for pellet/cladding mechanical interaction (PCMI) failure of light water reactor fuels under reactivity-initiated accident (RIA) conditions, this paper summarizes the recent important outcomes from research programs with the Nuclear Safety Research Reactor (NSRR). Applicability of current criteria, which are defined as a function of fuel burnup and possibility of introducing another parameter for new criteria were evaluated based on the results of the RIA-simulated pulse irradiation tests, post-test examinations, and supporting analytical work, such as the reevaluation of fuel enthalpies in earlier NSRR experiments. Failure-threshold curves based on cladding hydrogen content as a primary measure of fuel degradation have been proposed as a possible alternative that can be used to judge the occurrence of PCMI failure to ensure conservativeness in a more pertinent manner.     

Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions

Hydride precipitation along the radial-axial plane increases in high burn-up boiling water reactor (BWR) fuel claddings. The radially-oriented hydrides may have an important role during fuel behavior in a reactivity-initiated accident and may reduce ductility of the cladding under pellet-cladding mechanical interaction (PCMI) conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large rupture openings with an axial crack at room temperature and 373 K. The crack penetrated through cladding wall preferentially along the radial hydrides, and radial cross section showed cladding failure in a brittle manner. However, reduction in residual hoop strain by precipitation of the radial hydrides was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.     

Application results of a prototype ultrasonic liquid film sensor to a 7 MPa steam–water two-phase flow experiment

A prototype ultrasonic liquid film sensor was applied to a high-temperature steam–water two-phase flow experiment. The liquid film sensor was vertically installed in a loop which was connected to HUSTLE, a multi-purpose steam source test facility. The hydraulic diameter of the measurement section was 9.4 mm. The output waveforms of the sensor were acquired with a digital oscilloscope. The fluid temperature and system pressure were kept at 288 °C and 7.2 MPa, respectively, during the experiment. The pulse-echo method was used to calculate the liquid film thickness. The cross-correlation calculation was utilized to determine the time difference between the pulse reflected at the sensor surface and the pulse reflected at the liquid film surface. The time-averaged liquid film thicknesses were less than 0.055 mm in the annular flow condition. The increase of the time-averaged thickness was small with the change of the gas momentum flux. The film thicknesses measured with the sensor were compared with the past experimental results; the former were smaller than one-fourth of the thickness estimated as the mean film thickness. The comparison results suggested that the continuous liquid sublayer thickness was measured with the liquid film sensor.     

Modern Technology Applied in Advanced BWR (ABWR)

The Advanced Boiling Water Reactor (ABWR) has been developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced radiation exposure and radwaste. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. The ABWR incorporates the best proven features from BWR designs in Japan, the United States and Europe. The many new features are seen to provide superiority in terms of performance characteristics and economics relative to current LWR designs. The Tokyo Electric Power Co., Inc. recently announced the selection of General Electric Co., Hitachi, Ltd. and Toshiba Corp. to design and construct two lead Advanced Boiling Water Reactors as Unit 6 and 7 at the Kashiwazaki Kariwa Nuclear Power Station. Construction is scheduled for the early 1990's, and commercial operation planned for 1996 for Unit 6 and 1998 for Unit 7.