Boiling Water Reactor Pressure Vessel Research Articles

Study on structural integrity of reactor pressure vessel with various ablation under core meltdown accident

In-vessel retention is an important measure to alleviate the core meltdown accident in the third generation nuclear power plant. External Reactor Vessel Cooling(ERVC) is the key to the successful implementation of IVR. Due to the ablation caused by the molten pool in the lower head, the reactor pressure vessel(RPV) has various wall ablation and thinning, which poses a great threat to the integrity of the pressure boundary of the RPV. Based on the boiling water RPV as the research object, the model is established for the RPV with various ablation and local thinning by finite element method(FEM). The stress/strain response of the structure is calculated and analyzed under the core meltdown accident. The plastic and creep damage are assessed for the RPV in terms of its evolution and distribution with continuum damage mechanics(CDM). Due to the existence of internal pressure within the RPV, the effect of pressure on creep and plastic rupture is also studied within the required 72 h. According to the criterion of ductile exhaustion, the results show that the increase of internal pressure further reduces the load-bearing capacity of RPV.

Brittle fracture initiation in decommissioned boiling water reactor pressure vessel head weld

The brittle fracture initiation behavior and fracture toughness in the ductile-to-brittle transition region for a thermally-aged weld metal of a decommissioned reactor pressure vessel head (in operation at 288 °C for 23 effective full power years) and in the non-aged reference condition were investigated. The results show that brittle fracture initiated primarily from non-metallic inclusions. The correlation between fracture toughness and brittle fracture initiation type (inclusion debonding or breakage), initiator size, initiation location (as-welded or re-heated regions in the weld metal) were analysed. Despite that thermal ageing does not affect significantly the fracture toughness, it could promote the debonding as a brittle fracture primary initiation type. The influence of debonded inclusion on the evolution of cumulative damage and brittle fracture initiation was assessed using crystal plasticity modelling.

Effect of weld microstructure on brittle fracture initiation in the thermally-aged boiling water reactor pressure vessel head weld metal

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant. As-welded and reheated regions mainly consist of acicular and polygonal ferrite, respectively. Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions (0.5–2.5 µm) at the brittle fracture primary initiation sites. High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch. The density, geometry, and chemical composition of the primary initiation inclusions were investigated. A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the de-bonded interfaces between the uncracked inclusions and weld metal matrix. Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

Demonstration of Structural Integrity of Boiling Water Reactor Pressure Vessels Under Ultimate Response Guideline Operation

In recent years, the compound beyond-design-basis accident (BDBA), which combines earthquake, tsunami, or some other severe events to impact a nuclear power plant (NPP), has received more attention. After the Fukushima nuclear disaster, the licensee of NPPs in Taiwan established the ultimate response guideline (URG) that instructs operators to perform reactor depressurization, low-pressure water injection, and containment venting to prevent core meltdown and hydrogen explosion once long-term loss-of-power and water-supply events occur. In this paper, we employed the probabilistic fracture mechanics (PFM) method to evaluate the structural integrity of boiling water reactor (BWR) pressure vessels under URG operation. At first, models of the beltline shell welds for BWR vessels associated with the Pressure Vessel Research Users Facility-Exponential flaw distribution were built for the PFM Fracture Analysis of Vessels–Oak Ridge (FAVOR) code. Then, the thermal-hydraulic data of URG transients for Taiwan domestic BWRs were imposed as the loading conditions. The analysis results demonstrate that performing URG operation will not cause significant fracture probability even at extreme embrittlement conditions. If long-term station blackout occurs due to a compound BDBA, the URG procedures can prevent core damage and hydrogen explosion, while maintaining the structural integrity of the reactor pressure vessels.

Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event

This paper performs the probabilistic fracture analyses for boiling water reactor pressure vessels subjected to the low temperature over-pressure event. Using the FAVOR code, which was developed by the Oak Ridge National Laboratory in America, a conservative model with the plant specific data of domestic BWRs in Taiwan to the USNRC requirements is built. A hypothetical transient of low temperature over-pressure event which could severely challenge the integrity of reactor pressure vessel of BWR is assumed as the loading condition for the probabilistic fracture mechanics analyses. The effects of performing inservice inspection on the probability of failure are also simulated. The computed low probability of failure indicates that the analyzed reactor pressure vessel can maintain sufficient reliability even without performing any inservice inspections for the circumferential welds. It also indicates that extensive inservice inspections on shell welds cannot promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on reactor pressure vessel degradation. The developed probabilistic fracture mechanics model is helpful for the subsequent BWR regulation in Taiwan.

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed ...

A Modeling of BWR-MOX Assemblies Based on the Characteristics Method Combined with Advanced Self-Shielding Models

Calculations based on the characteristics method and different self-shielding models are presented for 9 × 9 boiling water reactor (BWR) assemblies fully loaded with mixed-oxide (MOX) fuel. The geometry of these assemblies was recovered from the BASALA experimental program. We have focused our study on three configurations simulating the different voiding conditions that an assembly can undergo in a BWR pressure vessel. A parametric study was carried out with respect to the spatial discretization, the tracking parameters, and the anisotropy order. Comparisons with Monte Carlo calculations in terms of keff, radiative capture, and fission rates were performed to validate the computational tools. The results are in good agreement between the stochastic and deterministic approaches. The mutual self-shielding model recently introduced within the framework of the Ribon extending self-shielding method appears to be useful for this type of assemblies. Indeed, in the calculation of these MOX benchmarks, the overlapping of resonances, especially between 238U and 240Pu, plays an important role due to the spectral strengthening of the flux as the voiding percentage is increased. The method of characteristics is shown to be adequate to perform accurate calculations handling a fine spatial discretization.

Flaw density examinations of a clad boiling water reactor pressure vessel segment : 38030 Cook, K.V.; McClung, R.W. Oak Ridge National Laboratory, Oak Ridge, Tennessee (United) States), NUREG-CR-4860, 30 pp. (Apr. 1987)

Flaw density examinations of a clad boiling water reactor pressure vessel segment : 38030 Cook, K.V.; McClung, R.W. Oak Ridge National Laboratory, Oak Ridge, Tennessee (United) States), NUREG-CR-4860, 30 pp. (Apr. 1987)

Flaw density examinations of a clad boiling water reactor pressure vessel segment : Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States), NUREG-CR-4860, 30 pp. (Apr. 1987)

Flaw density examinations of a clad boiling water reactor pressure vessel segment : Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States), NUREG-CR-4860, 30 pp. (Apr. 1987)

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Abstract Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.

Flaw density examinations of a clad boiling water reactor pressure vessel segment

As part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel. This boiling water reactor vessel was never in service. One objective was to evaluate the approximate 0.7- by 3-m (2- by 10-ft) segment of the vessel provided using ultrasonic flaw detection methods performed with both ASME Code techniques and supplemental ultrasonic methods. A second objective was to evaluate the inner surface stainless steel cladding for cracks with a high sensitivity penetrant examination. Both objectives were successfully completed. Five Code-recordable indications were detected ultrasonically; however, all were found to be anomalies associated with the cladding. One flaw was detected by the supplemental ultrasonic tests, and it was analyzed destructively. This flaw was pipelike indication, about 20 mm (0.8 in.) long extending along the length of the longitudinal weld in which it was located and was about 20 mm below the cladding surface. The flaw had a through-wall dimension (or length) of about 6 mm (0.24 in.) for an approximate 3-mm (0.1-in.) distance along the 20-mm major length. No flaws were detected by the penetrant examination of the cladding surface.

Experimental study on the application of acoustic emission during the hydraulic pressurization of a reactor pressure vessel

The investigations were aimed at demonstrating the state of the art of acoustic emission testing (AET) of reactor pressure vessels. The object under investigation was the large reactor pressure vessel of the MPA in Stuttgart, a boiling-water reactor pressure vessel, which was provided with a multitude of flaws in weld seams and in the base material. Six hydrostatic tests approximately up to the working pressure of a boiling-water reactor (71 bar) were carried out. In addition to the global multichannel locating technique, also local monitoring techniques were applied. Global location permitted a large number of different indications to be detected simultaneously. Not all of the known flaws did, however, show the expected number of AE events. On the other hand, it was possible to detect flaws previously unknown to the AE staff in some weld seams; these indications were confirmed by nondestructive testing. It was demonstrated that the locating accuracy of local monitoring using signal analysis was improved by a factor of 20 to 30 compared to global monitoring.

Nondestructive inspection of nozzle inner corner at boiling water reactor pressure vessels with ultrasound-flaw detection and crack depth determination

Abstract The demonstration of inservice inspection methods with ultrasound for the nozzle inner corner at boiling water reactor pressure vessels has shown that a detectability of cracks with a depth in the range of 5–10 mm is possible if optimal inspection parameters are chosen. The investigations concerning the choice of the optimal parameters have delivered the following results: 1. (1) Probe parameters: The angle of refraction should be in the range of 70° and a probe frequency of 1 MHz should be used. The tilting angle can be chose between 18 and 22°. 2. (2) Sensitivity: The calibration of the sensitivity can be carried out with the help of a 7 mm side-drilled hole in a distance corresponding to the maximum sound path for the inspection. 3. (3) Scanning range of the probe: The probe shall be moved in a range of about 100 mm in order to cover the whole gap between minimal and maximal radial distance for an elliptical scanning. 4. (4) The evaluation of indications can be done based on C -scan presentations. In doubtful cases the time-of-flight curves will deliver some useful information. The first hint on the defect size is delivered by the extension of the indication in a C -scan presentation. A dynamic extension can be evaluated and correlated to the real defect size if sufficient knowledge about the specific relationship is available. For critical findings a correct flaw sizing cannot be avoided. In those circumstances it seems to be necessary to apply techniques like the acoustic holography or others.

Radiation Effects in Boiling Water Reactor Pressure Vessel Steels

Effect of neutron irradiation on notch toughness properties of Boiling Water Reactor pressure vessel steels was determined. Samples from several heats of plate, weld metal, and forgings were irradiated to three different fluence levels and tested. A statistical evaluation of the data was conducted to determine regression analysis mean decreases in upper shelf energy and increases in transition temperature versus fluence.

Ultrasonic inspection of the spherically shaped perforated areas of light water reactor pressure vessels

The development is described of an ultrasonic system which will test the areas close to the cladded inner surface between the nozzles in the bottom of boiling water reactor pressure vessels and at the top of pressurized water reactor pressure vessels. The technique uses the angle or corner effect in a single probe operation from the outside. A main problem found with this method was the strong back scattering of ultrasonic energy from the austenitic cladding. By varying the frequency, crystal area and angle of incidence used, optimum values for these parameters have been determined for this application.