Head Of Reactor Pressure Vessel Research Articles

Microscopic characterisation of brittle fracture initiation in irradiated and thermally aged low-alloy steel welds of a decommissioned reactor pressure vessel

Microstructure has a significant effect on material's integrity and in a heterogeneous weld microstructure the discontinuities affect the brittle fracture initiation and propagation and determine the fracture toughness. The knowledge of brittle fracture initiation mechanisms in high-Mn/high-Ni welds is limited. The brittle fracture initiation behaviour of the decommissioned Barsebäck Unit 2 reactor pressure vessel (RPV) welds of high-Mn/high-Ni weld metal from three different locations, the RPV head and the beltline regions, were investigated and compared with specimens from the surveillance program with high fluence. Systematic fractography has been performed on impact and fracture toughness specimens and the main features of the brittle fracture initiation in the component weld are presented and discussed. Two main types of initiators are identified as the weakest links to initiate the cleavage fracture and the initiation mechanism is found independent from the operation condition. The high-fluence surveillance specimens have a larger amount of intergranular cracking. The cleavage fracture initiation appears to be independent of the operation conditions but dependent on the welding process and metallurgical features. The findings aid in the development of improved material-property correlations which will result in better computational tools for predicting aging of welds based on microstructure.

On comparison of fracture toughness of irradiated and thermally aged decommissioned nuclear weld metals with miniature specimen test technique

This paper investigates the fracture toughness of thermally aged and irradiated weld metals extracted from head and beltline regions of a decommissioned nuclear reactor pressure vessel. Miniature compact tension specimens are fabricated from the thermally aged reactor pressure vessel head weld as well from the circumferential and axial beltline welds having fluences of 2.90 × 1016n/cm2 and 7.94 × 1017n/cm2, respectively. Master curve approach in accordance with the ASTM E1921 is applied to determine the reference temperature T0 and it is found to be − 113.5 °C, −104.4 °C, and − 89.3 °C for the reactor pressure vessel head, axial beltline, and circumferential beltline welds, respectively. In addition, the multimodal reference temperature Tm to assess the homogeneity of welds as well as key curve analysis for the quality assurance of testing are also provided. Fractographic analysis identified variations in crack initiation sites and microstructural inhomogeneities, particularly in the irradiated beltline welds. Overall, fracture mechanical testing of the materials demonstrated that miniature compact tension specimens are effective for characterizing the fracture toughness of limited surveillance weld materials, revealing differences due to irradiation and weld location.

Evaluation of RPV closure head penetration integrity considering fitting and residual stresses

To ensure the structure integrity of the Reactor Pressure Vessel (RPV), the main challenge is the embrittlement of beltline material. However, the stress is more complex for the regions of Closure Head Penetration (CHP). Feedback of leakage at the CHP has been found during in-service inspection. Although the detailed mechanical evaluation of the RPV head and the CHP has been carried out by many international institutions, there are limited studies to concern structure integrity of CHP tube. The systematic analysis of the influence of material strength matching, assemble stress, the Welding Residual (WR) stress and work pressure load on the fracture is also limited. In this paper, the systematic influence of these factors is analyzed, and the distribution characteristics of Stress Intensity Factor (SIF) and the crack propagation characteristics are studied. The 2D model and accurate 3D model are carried out using the finite element method. The results show that the WR stress on the outer surface of the CHP and the area adjacent to the weld is larger than the yield strength of the material. There is a gap between the RPV and the upper part of the CHP. The attenuation rate of the WR stress in the outlet direction of RPV is lower than that in the inlet direction, so cracks are more likely to occur in the out of direction. The influence of WR stress on the circumferential stress distribution and fracture parameters in the wall thickness direction of CHP is bigger than the influence of assemble load and design condition load. However, the overall stress of CHP is low. The SIF values along the crack front have an asymmetric distribution, and the SIF value is very small. As the fatigue crack propagation life is much longer than the assumed limit in the design lifetime, the risk of fracture and crack fatigue propagation caused by mechanical stress is low.

CNN-based automated approach to crack-feature detection in steam cycle components

Periodic manual inspection by trained specialists is an important element of asset management in the nuclear industry. Detection of cracks caused by stress corrosion is an important element of remote visual inspection (RVI) in power plant steam generator components such as boilers, superheaters and reheaters. Challenges exist in the interpretation of RVI footage, such as high degree of concentration for reviewing lengthy and disorienting footage due to narrow field of view offered by endoscope. Deep learning is considered useful to automate crack detection process for improved efficiency and accuracy, and has enjoyed success in related applications. This article utilises a new application of automated crack feature detection in steam cycle components to demonstrate a transferrable data-driven framework for a variety of anomaly inspections in such structures. Specifically, a case study of superheater (a type of reactor pressure vessel head) anomaly inspection is presented to automatically detect regions of crack-like features in inspection footage with a good accuracy of 92.97 % using convolutional neural network (CNN), even in challenging cases. Due to the black-box nature of the CNN classification, the explicability of the classification results is discussed to enhance the trustworthiness of the detection system.

Structural integrity assessment of a closure head of a PWR reactor pressure vessel under pressurized thermal shock

The pressurized thermal shock induced during the accidental conditions is the potential threat to the structural integrity of reactor pressure vessel. The pressurized thermal shock puts the reactor pressure vessel under detrimental thermo-mechanical stresses, and a small crack on the surface, particularly at highly stressed locations, may initiate, propagate, become critical, and eventually fail the component. Therefore, critical components are needed to be assessed. For structural integrity analysis of class-I components under pressurized thermal shock, a methodology is proposed in this research article. As a case study, the structural integrity assessment of the closure head (class-I component) of the RPV under pressurized thermal shock is presented here.

A scoping study on remelting process of a debris bed in the lower head of reactor pressure vessel

Coolability and retention of core melt (corium) in the lower head of reactor pressure vessel (RPV) has been accepted as a severe accident management strategy to maintain the reactor pressure vessel (RPV) integrity of a light water reactor. To qualify the in-vessel melt retention strategy, lots of studies have been performed to investigate the natural convection heat transfer of a melt pool in the lower head. However, little work has been attributed to the precursory phase of the melt pool, i.e., the remelting process of a debris bed which is formed in the lower head at the very beginning of corium relocation from the core the lower head. The present study is motivated to conduct an experimental study on the debris remelting process. For this purpose, a dedicated test facility named COREM (COrium REMelting) is conceived and constructed, which features internal heating of electromagnetic induction and visualization of debris remelting dynamics. The test section is a semicircular vessel representing a slice of scaled-down RPV lower head, whose front and back walls are made of transparent tempered glass which facilitate visualization and induction heating of the debris bed. Fiber probes with multiple optical temperature sensors are mounted in the semicircular wall of the test section to measure its temperature distribution. In the scoping test, n-octanol and Wood's metal are selected as the simulant materials of metallic and oxidic components of corium, respectively, and their particles ware loaded in the test section to form a debris bed. This paper presents the first two scoping tests which have been performed with the COREM facility so far, under different mixing ratios of two debris materials. The measured data include the photography of debris remelting processes and the temperatures in the debris bed and the semicircular wall. Based on the temperature distributions, heat flux profile along the semicircular vessel wall is also estimated. The scoping tests well reproduce the dynamic process of debris remelting, with two distinct stages of fusion of n-octanol and Wood’s metal, i.e., melting successively from low to high melting-point debris particles. During the remelting process, the vessel wall temperatures increase with the polar angle firstly and then decrease gradually, leading to the highest temperature appearing in the middle of the lower layer of the stratified molten pool which is finally formed.

Fracture mechanics analysis of a closure head of a PWR reactor pressure vessel the effect of bolt pre-load and thermal stresses

The effect of the bolt pre-load and thermal stresses in addition to pressure stresses on the structural integrity of a closure head of an RPV is presented here. The trends of the stress intensity factors (SIFs) of the surface cracks have been investigated under the combined effects of the loading conditions. A contradictory behavior of the surface cracks at the inside and outside locations of the head is observed due to thermal stresses, while the bolt pre-load decreases the integrity of the head at both locations. The maximum SIFs of the analyzed cracks are also found well below the material’s fracture toughness.

Study on the influences of RPV deformation on CHF under IVR conditions

After the reactor core melt, the thermal impact of high temperature molten mixture to the lower head of reactor pressure vessel (RPV) will cause the shape change of cooling channel constituted by reactor pressure vessel outer surface and other supporting structures. This phenomenon may lead to local heat transfer deterioration and then cause the failure of in-vessel retention (IVR). Therefore, it is necessary to study the reactor pressure vessel deformation under in-vessel retention conditions. This paper proposes a numerical method for the CHF of the outer wall of reactor pressure vessel using ANSYS and FLUENT. The geometry of coolant channel is modeled by ANSYS with the coupling of temperature and mechanical field analysis in ANSYS, and the critical heat flux (CHF) of reactor pressure vessel outer surface is further predicted by FLUENT. With the comparing between test data and calculation, the calculation methods are verified. The results show that the CHF will be decreased by the deformation of PRV caused by the core melt.

Fracture mechanics analysis of a closure head of a PWR reactor pressure vessel LEFM-based SIF

In this study, fracture mechanics analysis of a closure head of a PWR reactor pressure vessel has been performed under accidental conditions. For the purpose, a wide range of elliptical surface cracks starting from minute cracks to worst-case cracks was postulated at the highest stress concentration locations of the head. The analyzed accidental conditions are small break loss of coolant accident (SB-LOCA) and Rancho-Seco transient (RST). The maximum linear elastic fracture mechanics-based stress intensity factors of the worst-case crack ‘a = 0.25 th; a = 0.33 c’ under SB-LOCA and RST are 104.37 MPa√m and 137.12 MPa√m respectively, which are less than the fracture toughness value of the material i.e. 155.6 MPa√m. Hence, the accidental conditions do not threaten the structural integrity of the proposed closure head.

An investigation of characteristics of flow and heat transfer in the two-layer molten pool under swing motion condition

In-Vessel Retention (IVR) of molten corium is an important strategy for management of severe accidents adopted by some active service generation II plus reactors and advanced generation III reactors like AP1000, Hualong one and APR1400. One of the most important factors contributing to IVR analysis can be attributed to the characteristics of heat transfer in the molten pool in the lower head of reactor pressure vessel (RPV), which directly determines the heat load imposed on the vessel wall of the RPV lower head. Besides, IVR is also an important strategy for management of severe accidents of Ocean Floating Reactor (OFR), whose flow and heat transfer characteristics are influenced by the ocean motion conditions. However, since all previous molten pool experiments were conducted under static condition, characteristics of the heat transfer in the molten pool thus obtained may be different from those under ocean motion conditions, thus not applicable to the latter. To address this issue, in this study, an experimental system was established, which includes a two-layer molten pool test facility and a swing motion platform. Based upon the system, the effects of the swing angle and swing period on the characteristics of heat transfer in the two-layer melt pool were investigated. In addition, a scaling method was proposed to select simulant materials for the two-layer corium pool. Results of the experiment show that the swing period and angle have a significant effect on the temperature field of the molten pool and the rate of heat transfer. We thus suggest that, in future studies, new heat transfer correlations, models and IVR analysis method for ocean floating reactors should be researched and developed to evaluate the IVR design of ocean floating reactors, which is critical for management of severe accidents of the ocean floating reactors.

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.

Numerical investigation of fluid–solid interaction during debris bed formation based on MPS-DEM

The core may melt down and a debris bed with specific shape may be formed on the lower head of Reactor Pressure Vessel (RPV). The solid-to-solid collision has not been concerned thoroughly in the previous studies. In order to solve the problems of fluid–solid interaction and solid collision, the MPS-DEM code is developed based on Moving Particle Semi-implicit (MPS) method and Discrete Element Method (DEM). In this paper, the MPS models, DEM models are introduced, and the coupling method is validated by particle flow and collapse tests. Besides, the two-dimensional (2D) and three-dimensional (3D) simulations of the QUEOS-Q21 test are performed. The results of numerical simulation are in accordance with that of experiments, which indicates that the MPS-DEM method has the potential to be applied to study the movement characteristics of the debris particles during the debris bed formation without heat transfer, and has application prospect in the severe accident analysis.

Vessel Failure Analysis of a Boiling Water Reactor During a Severe Accident

In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV’s structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress σ at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented.

A Scoping Study on Remelting Process of a Debris Bed in the Lower Head of Reactor Pressure Vessel

A Scoping Study on Remelting Process of a Debris Bed in the Lower Head of Reactor Pressure Vessel

Experimental research of heat transfer in heavy metallic layer under different top boundary conditions

The External Reactor Vessel Cooling (ERVC) is crucial to realizing In-Vessel Retention (IVR) strategy when reactor core degradation happens; consequently, fission products would be limited in Reactor Pressure Vessel (RPV). Because of the fission products in the lower head of RPV, vessel integrity is threatened. Due to the immiscibility of oxide and metal materials, the molten pool in the lower head of RPV would turn to a stratified configuration. Two-layer (light metallic layer atop of oxidic layer) and three-layer (light metallic layer, oxidic layer, heavy metallic layer) configurations have been proposed. The possible heavy metallic layer has a different shape, top boundary conditions, and decay heat compared with the other two layers. Thereby, the objective of the heavy metallic layer experiment is to study its heat transfer characteristics, and two possible top boundary conditions are performed. The heating cables simulating the decay heat are placed in a hemispheric test section (diameter = 2.4 m). Water is used as the experimental simulant with a height of 0.3 m. The experimental results demonstrate that increasing the top cooling rate, the ratio of top cooling power to sidewall cooling power, and the ratio of upward heat flux to sideward heat flux increase. In addition, they share almost the same trend. The central melt temperature rises with the height increment, and the maximum normalized temperature reaches about 1.3 for both top cooling and insulated boundary conditions. Thermal stratification is found in the experiment, and the possible natural convection is limited in the heavy metallic layer. Correlations of UCLA, Mini-ACOPO, and RASPLAV reasonably predict the heavy metallic layer's normalized sideward heat flux distribution.

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.

Numerical Simulation and Validation for Early Core Degradation Phase under Severe Accidents

Early core degradation determines the amount of hydrogen generated by cladding oxidation as well as the temperature, the mass, and the composition of corium that further relocates into the lower head of reactor pressure vessel (RPV), which is essential for the effectiveness analysis of in-vessel retention (IVR) and hydrogen recombiners. In this paper, the mechanisms of controlling phenomena in the early phase of core degradation are analysed at first. Then, numerical models adopted to calculate (1) core heating up, (2) cladding oxidation, (3) dissolution between molten zirconium and fuel pellets, and (4) formation of a molten pool in the core active section are presented. Compared with integral codes for severe accident analysis (such as MAAP and MELCOR), the models in this paper are established at the fuel pin level and the calculation is performed in 3D, which can capture the detail local phenomena during the core degradation and eliminate the average effect due to equivalent rings used in integral codes. In addition, most of the control equations in this paper are calculated by implicit schemes, which can improve the accuracy and stability of the calculation. In the simulation, the calculation oxidation is calculated by using the oxygen diffusion model, while the dissolution is calculated with Kim, Hayward, Hofmann, and IBRAE models to perform uncertainty analysis. For the validation, the cladding oxidation model is verified by Olander theoretical cases in the conditions of both steam-rich and steam-starved. The dissolution models are validated by the RIAR experiment. The code is overall verified by Phebus FPT0 on the integral phase of core early degradation. According to the simulation results, it can be inferred that the dissolution reaction between the molten zirconium and fuel pellets is the main reason for the melting of UO2 at low temperature. In the case of starved steam, part of the fuel pellets can melt down even at 2248 K and relocate to the bottom of the core, which is much lower than the melting point of UO2 (3113 K).

Prediction of critical heat flux on surface under boiling in case of IVR-ERVC by using CFD simulation

The efficacy of in-vessel through external reactor vessel cooling (IVR-ERVC) is strongly correlated with the critical heat flux (CHF) on the lower head of reactor pressure vessel (RPV). The high limit of CHF prevents the RPV from failure by extracting much heat out of the vessel through the circulation of working fluid in external flow channel. In present study, the deionized water is allowed to flow through curved flow channel having radius of curvature of 368mm and hydraulic diameter of 40mm with square cross-section. The heat flux of 1500kWm−2 is applied perpendicular to the convex wall of the channel. The CHF on the wall of flow channel has been predicted by incorporating CFD based two-phase (liquid-vapor) boiling model. The investigations on spatial variation of pressure, temperature, and velocity and heat flux acquired as a result of numerical simulations have been discussed in detail. The CHF and average HTC is predicted as 1.798e+04kWm−2 and 38.981kWm−2K−1 respectively.

An experimental study on natural circulation flow characteristics under passive IVR-ERVC conditions

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated by the molten debris relocated to the lower head of reactor pressure vessel needs to be removed continuously to prevent vessel failure.Besides the local critical heat flux (CHF) of outer wall surface which is of the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure, reactor pressure vessel external cooling II test facility (REPEC-II) was designed and constructed in Shanghai Jiao Tong University (SJTU). During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics with the experimental observation and measurement are investigated under IVR-ERVC conditions.Based on the abundant results acquired in the test campaign, it is attempt to evaluate the ERVC performances and trends under various practical engineered conditions in this paper. From this study, the following conclusions can be made: The influence of different power shapes on the two-phase natural circulation flow is not obvious, while the impact on the resistance characteristics of the outlet section is very significant. Among all the test conditions (power shapes, subcooling, flow path configuration and water level), the resistance characteristics of the outlet section change the most compared with other sections. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

Microscopic damage mechanism of SA508 Gr3 steel in ultra-high temperature creep

The lower head of reactor pressure vessel (RPV) will endure a great temperature gradient above the phase transition temperature, and the creep and fracture will be the primary failure mode for the RPV material in such a situation. The interrupted creep tests were performed on a typical RPV material, SA508 Gr3 steel, at 800 °C. The microstructure of different creep stages was examined by scanning electron microscopy and transmission electron microscopy. The results showed that the microscopic damage is mainly induced by creep cavities and coarse second-phase particles. Furthermore, the volume fractions of creep cavities and coarse second-phase particles show a linear relationship with the extended creep time. The second-phase particles are determined to be MoC in the second creep stage and Mo2C in the third creep stage, according to the results of selected-area electron diffraction pattern. Combined with energy-dispersive spectrum analysis, the segregation of precipitates caused by the migration of atoms is finally unveiled, which leads to the coarsening of the particles.