Integrity Of Reactor Pressure Vessel Research Articles

Numerical simulation of in-vessel steam explosion for China third-generation PWR

In this study, in order to evaluate the Reactor Pressure Vessel (RPV) integrity under in-vessel steam explosion loads for China third-generation Pressurized Water Reactor (PWR), 2D and 3D numerical simulation of in-vessel steam explosion is conducted using MC3D code. The sensitivity analyses for in-vessel steam explosion are performed, including break area, water temperature, water level, melt temperature, and break height. The results show that the premixing stage cannot cause RPV failure and the possibility of the steam explosion blowing the RPV upper head off is almost non-existent. When the break area increases, the maximum pressure and impulse of the RPV inner wall increase obviously first and then generally decrease slightly. With the increase of water temperature, the maximum pressure and impulse first increase and then decrease. When the water level is higher, the maximum pressure variation is small. When the water level is very low, the maximum pressure decreases. The maximum impulse generally decreases with the decrease of the water level. When the melt temperature exceeds 2850 K, the steam explosion is not sensitive to the melt temperature. With the increase of the break height, the maximum pressure and maximum impulse first increase and then decrease. Furthermore, based on the results of sensitivity analysis, two cases of conservative conditions are selected, and the RPV integrity under in-vessel steam explosion loads for China third-generation PWR is considered. The result shows that the integrity of the RPV can be maintained under the in-vessel steam explosion loads for China third-generation PWR with a large margin. These studies provide some reference to improve the severe accident management strategy in China.

CFD Applications to Pressurized Thermal Shock-Related Phenomena

In pressurized water reactor accident scenarios, the injection of water from the emergency core cooling system (ECCS) (ECC injection) might induce a pressurized thermal shock (PTS), affecting the reactor pressure vessel (RPV) integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal–hydraulic and structural analyses. The thermal–hydraulic approach is particularly crucial, and reliable computational fluid dynamic (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer (DC). Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD simulation can benefit from complement evaluations with thermal–hydraulic system analysis codes. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the Rossendorf coolant mixing model (ROCOM), transient two-phase flow (TOPFLOW), upper plenum test facility (UPTF), and large-scale test facility (LSTF). The results revealed that in single-phase flow CFD simulation, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using large eddy simulation (LES) have been reported. For two-phase flow analysis of air–water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k–ε and shear stress transport (SST) k–ω models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam–water conditions, many studies have used two-fluid models and Reynolds-averaged Navier-Stoke (RANS), and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation.

MOPSO-based structure optimization on RPV sealing performance with machine learning method

The structure integrity of reactor pressure vessels (RPV) plays an important role in the safety of nuclear power plants, which mainly concerns the risk of sealing failure. In the present work, a hybrid method consisting of an artificial neural network (ANN) and multi-objective particle swarm optimization (MOPSO) is proposed, aiming to find an optimized RPV structure with improved sealing performance. The ANN model provides a more accurate prediction of sealing performance than the radial basis function neural network (RBFNN) model. There is also a better generalization ability in the ANN model with R2 above 0.96 and RMSE below 0.06 in the test set. With ANN as a surrogate model, multi-objective optimization is applied in RPV structure design. Compared with that of the original RPV structure, the sealing performance can be improved by even over 25% in the optimized structure. The application of ANN surrogate models can save more than 99% computation cost, compared with optimization based on finite element analysis (FEA).

Enhancement of the critical heat flux for downward-facing saturated pool boiling on the reticular hollow shell structure surfaces

In-vessel retention (IVR) technology was currently an effective strategy for retaining molten materials in the reactor pressure vessel (RPV) when the nuclear power plants occurred severe accident, further ensuring the integrity of RPV. The effectiveness of IVR depended on the critical heat flux (CHF) of external reactor vessel cooling (ERVC). A novel type of reticular hollow shell structure (RHS) was designed to meet this need. Five RHS surfaces were designed and the pool boiling heat transfer performance were investigated in saturated deionized water, the experimental inclination angles were 5°, 30°, 45°, 60° and 90°. The boiling heat transfer coefficient (BHTC) and CHF were measured by steady-state heating and transient quenching methods. The experiment results showed that the CHF increased with the increase of inclination angles, and the CHF of all RHS surfaces were significantly enhanced compared with the plain surface. The maximum CHF value and the maximum CHF enhancement exhibited on surface 4, with a maximum CHF of 2992.2 kW/m2.When the inclination angle was 30°, the maximum CHF could reach 2338.6 kW/m2, which was 160.3% enhancement over the plain surface. Surface 1 and surface 3 had better BHTCs, and the maximum values were 178.3 kW/(m2·K) and 126.8 kW/(m2·K), respectively, which were much better than other structured surfaces. The RHS could form a gas–liquid conversion system, which was composed of cavities and alternating channels to increase the surface nucleation sitesand accelerate the circulation of cooling water, thus obtained the high CHF and high BHTC characteristics. RHS provided a large enough CHF margin to ensure that the RPV was continuously cooled and its integrity was maintained. Finally, the behavior of bubbles generation, coalescence, sliding and detachment in reticular hollow shell structure were systematically analyzed.

Creep behavior and life prediction of a reactor pressure vessel steel above phase‐transformation temperature via a deformation mechanism‐based creep model

AbstractFor nuclear power generation as a carbon‐neutral energy source, in‐vessel retention (IVR) must be implemented to maintain the structural integrity of nuclear reactor pressure vessel (RPV) for more than 72 h under severe accidental conditions. This technology requires accurate prediction of creep deformation and life of RPV material being operated under pressure and extremely high‐temperature gradient. The current work develops a simplified deformation‐mechanism‐based true‐stress (DMTS) model for creep behavior/life‐prediction of SA508 Gr.3 steel, a typical RPV material, above the phase transformation temperatures (800–1000°C). This model is used to evaluate the time to specific creep strain (t3% and t5%) and rupture (tr), in comparison with popular empirical methods such as Orr–Sherby–Dorn (OSD) and Larson–Miller (LM). The simplified DMTS model achieves an excellent agreement with the experimental observations. The controlling deformation mechanisms are also discussed by metallurgical examinations, which provide the physical premise for the model development and application.

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.

Influence analysis of alloy elements on irradiation embrittlement of RPV steel based on deep neural network

Reactor pressure vessel (RPV) is the most important core equipment in PWR. Its service life determines the service life of nuclear power plant and directly affects the economy and safety of nuclear power plant. Because RPV is serviced at high temperature, high pressure and high energy neutrons for a long time, the properties of RPV steel will significantly degrade, in which irradiation embrittlement is the most important factor for the structural integrity of RPV. In this work, about 700 groups of data such as composition, irradiation conditions and ductile brittle transition temperature of RPV steel are collected, and the data are cleaned and screened for modelling by machine learning. The deep neural network is used for establishing the correlation between key component and irradiation embrittlement of RPV steel. The results show that the lower flux of neutron irradiation will make the radiation embrittlement effect of RPV steel more obvious at the same neutron fluence. Cu, P and Ni are the key factors to influence the △DBTT of RPV steel. The synergistic effect of Cu and Ni on irradiation embrittlement is greater than that of Cu and Mn. These results will help to promote the optimization design of new RPV steel.

Study on the safety characteristics of molten corium in-vessel retention of annular fuel core

In the process of reactor severe accidents, maintaining the integrity of reactor pressure vessel to achieve IVR (In-Vessel Retention) of molten corium plays an important role in accidents mitigation and suppression. The ERVC (External Reactor Vessel Cooling) is an important technology to realize IVR. As a new economical annular fuel, the increase of zirconium content and power density in the core may threaten the integrity of lower head under severe accidents. In this paper, the IVRASA (In-Vessel Retention Analysis code in Severe Accident) is used to analyze the heat transfer characteristics of lower head molten pool corresponding to annular fuel core. The components of two-layer and three-layer configurations of molten pool are determined based on the geometric structures of annular fuel and verified heavy metal layer model. In terms of the integrity of lower head, the power density of the annular fuel core can be increased to about 110%.

Stress intensity factor assessment for reactor pressure vessel nozzles containing postulated corner cracks

Due to the structural discontinuity of geometry having higher stress concentration, the nozzles may dominate the structural integrity of reactor pressure vessel (RPV) system, even the material radiation embrittlement is insignificant. The nonmandatory Appendix G in Section XI of the ASME Boiler and Pressure Vessel code provides a fracture mechanics-based methodology using magnification factors to calculate the stress intensity factor of reference crack tips as long as the stress distribution through the nozzle section is available. This paper establishes various three-dimensional finite element models of Taiwan domestic RPV nozzles to conduct a series of stress and linear elastic fracture mechanics analyses to estimate the stress intensity factors. The circular and elliptical reference corner cracks with depths of one-fourth and one-tenth thickness of the vessel wall along the 45° path are considered. Two evaluation results using the ASME method and the finite element linear elastic fracture mechanics analysis are compared. It is found that the ASME method produces conservative values of stress intensity factor at crack tips of circular corner cracks but underestimates the elliptical corner cracks. Finally, an adjusted magnification factor formula is proposed to obtain more accurate stress intensity factors for elliptical-shaped nozzle corner cracks.

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.

Identifying Thermodynamic Mechanisms Affecting Reactor Pressure Vessel Integrity During Severe Nuclear Accidents Simulated by Laser Heating at the Laboratory Scale

In this work, laser heating is used to experimentally investigate the high-temperature behavior of the U-Fe-Zr-O system using arc-melted samples with various nominal compositions. Three-phase transitions are observed in the vicinity of ~1100, ~1700, and ~2200 K. Principal component analysis of the phase transition temperatures in the course of laser-heating thermal cycling indicates that the phase transition around ~1100 K is driven by the interaction of stainless steel (SS) with metallic U, the phase transition around ~1700 K by the melting of stainless steel, and the phase transition above ~2000 K by the eutectic melting of UO2. The results also reveal two hitherto overlooked interactions in the U-Fe-Zr-O system, which could have severe consequences for the containment of corium inside the reactor pressure vessel (RPV). First, the phase transition temperatures of the samples varied extensively as a result of the laser-driven rapid thermal cycling. Variations of up to 390 K were observed in the phase transition temperatures, suggesting that depending on the initial conditions of corium formation, the corium-driven ablation of the RPV wall could commence significantly earlier than the current state-of-the-art severe accident codes would predict. Additionally, evidence of a large exothermic reaction between zirconium and molten steel was observed upon SS melting. Such phenomenon may also be driven by material segregation during fast heating and cooling. If such a mechanism is activated during a severe nuclear accident, it can have an important impact on the overall thermal balance of the RPV.

Remarkable embrittlement and its origins in quenched reactor pressure vessel steels

Reactor pressure vessels (RPVs) are one of the main barriers against nuclear accidents, therefore good toughness is indispensable for keeping the integrity of RPVs to prevent the leakage of radioactive substances. It is well known that high temperature and large thermal gradient of nuclear accidents would lead to embrittlement or even rupture of RPVs. However, how damages originate in severe nuclear accidents is still indistinct. This study is aimed to reveal the embrittlement mechanism under a large temperature gradient in quenched RPV steels. It is found that certain strengthening as well as remarkable embrittlement occurs after quenched, further investigations find it probably derive from the generation of plenty of Moiré fringes and dislocation networks, which are supposed to be generated by shearing of crystal planes. This paper reports the deterioration phenomenon caused by a large thermal gradient, reveals the origins remarkable embrittlement and indicates that extra attention must be aroused to take this thermal-stress-induced embrittlement into consideration in nuclear accidents analysis.

Investigation on three-layer coupled molten pool heat transfer experiment with three different materials

In the core meltdown accident, fission products are limited in the vessel bottom as indicated in the In-Vessel Retention (IVR) strategy analyses, resulting in forming a three-layer configuration that would threaten the integrity of Reactor Pressure Vessel (RPV). The three-layer test has been carried out with the Three layeR cOupled molten pool heat tranSfer Experiment (TROSE) apparatus to investigate the heat transfer in this three-layer configuration. It uses mineral oil (height = 0.2 m), water (height = 0.7 m), and Cerrobend alloy (height = 0.3 m) as the simulants for the top, middle, and bottom layers, individually. TROSE test section is a three-dimension hemispheric facility whose diameter and thickness are 2.4 m and 0.025 m, respectively. Since the sidewall temperature is smaller than the melting point of Cerrobend alloy (70 °C), the crust is formed in the bottom layer, affecting its dominant heat transfer mode. The experimental results demonstrate that upward heat transfer is found at the interface of the middle and bottom layers. Besides, the sideward heat flux decreases with the increase of the polar angle. Furthermore, heat transfer correlations of Churchill-Chu, mini-ACOPO, and COPRA could reasonably predict the experimental ratio of top cooling power to sidewall cooling power (Pup/Pdn). Finally, the previous correlations of the sideward heat flux and Nusselt number distributions, such as Mini-ACOPO correlation, can reasonably predict the experimental results of the middle layer; however, they are not suitable for the top and bottom layers.

Structural integrity of reactor pressure vessels (1)

Structural integrity of reactor pressure vessels (1)

Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock

This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.

A Versatile Methodology for Reactor Pressure Vessel Aging Assessments

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime threshold. From the perspective of possible further lifetime extension, satisfying safety requirements is a priority. Characterization of the structural integrity of the reactor pressure vessel (RPV) is an important issue as it is a guiding parameter that influences the reactor lifetime. Embrittlement of RPV material is primarily induced by the bombardment of fast neutrons (with energies greater than 1 MeV). Consequently, fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of RPV. However, future RPV aging assessments might lean on new variables with respect to current laws, such as neutron fluence considering the whole neutron spectrum or displacements per atom (dpa) since the latter is more representative of overall damage generated in the RPV. In order to meet these challenges, a versatile calculation scheme for RPV aging assessments is proposed in this paper. The developed methodology allows one to compute (fast and non-fast) neutron fluence as well as dpa rate, using the Norgett-Robinson-Torrens dpa model and the Athermal Recombination Corrected dpa model, for a wide azimuthal and axial range on the RPV and in the capsules of the aging monitoring program (which contain dosimeters and vessel material samples). This methodology is based on a coupling between deterministic (CASMO5 and SIMULATE5) and Monte Carlo (MCNP6) numerical approaches. First, the deterministic approach is used to evaluate the full-core fission neutron source term. Second, Monte Carlo modeling is used to perform the neutron attenuation from the core to sites of interest, such as the RPV. The computational efficiency, accuracy, and potential benefits of the methodology are presented. Moreover, the frequency at which neutron transport calculations should be performed in order to obtain sufficiently accurate time-integrated data over a reactor cycle is discussed. Finally, the validity of the fast neutron fluence as an indicator of RPV aging is compared against the use of dpa.

Mechanistic critical heat flux prediction for in-vessel retention conditions

In-vessel retention (IVR) is considered a feasible technique to keep reactor pressure vessel (RPV) integrity in a severe reactor accident. For light water reactor (LWR), the effectiveness of this strategy relies soundly on the critical heat flux (CHF) distribution over the external surface of the lower plenum of RPV, whose orientation varies gradually from downward-facing horizontal to vertical. The CHF prediction capability of the liquid sublayer dryout model is efficient for high mass flux in vertical flow boiling conditions. This paper focuses on how to adapt the model to the changed orientation of the heating surface. Bubble departure diameter (dB) and net vapor generation point (NVG), the starting point for the void fraction developing in a heating channel, is one of the important key points in the CHF prediction. Therefore, to assess the predictive potential of CHF under IVR, experimental research was performed to measure bubble departure diameter and NVG for a changing heating surface orientation from downward-facing horizontal to vertical with a mechanistic model basing on the force balance. A modified liquid sublayer dryout model was then proposed where the channel orientation effect is considered to measure the bubble departure diameter (vapor blanket diameter) using the improved force balance model. The NVG is modified according to the departure diameter. The predicted departure diameter and subcooling at NVG show good consistency with the experimental data, and the modified liquid sublayer dryout model can predict the CHF data with an average relative error of 18.36% in IVR.

Experimental study and modeling of bubble lift-off diameter in subcooled flow boiling including the inclination effect of the heating surface

ABSTRACT In-vessel retention (IVR) is a viable method of preserving reactor pressure vessel integrity in severe reactor accidents. Accurate knowledge of the bubble lift-off diameter for a downward-facing inclined heating surface is vital for successfully implementing IVR following the critical heat flux (CHF). A research facility was designed to visualize bubble lift-off for different heating surface orientations from downward-facing horizontal to vertical in IVR and experimentally determined the bubble lift-off diameter. A significant effect of inclination angle on lift-off diameter was found. The maximum lift-off diameter was achieved in the downward-facing horizontal position of the heating surface, whereas the minimum one was with vertical orientation. The existing lift-off models of Sugrue et al. and Situ et al. follow the experimental data parametrically, but with significant average errors of 51.99% and 102.08%, respectively. These models either ignore bubble sliding velocity or assume it is constant with respect to the local liquid velocity. An improved mechanistic model is proposed considering the bubble velocity as a dynamic variable for the lift-off diameter including the inclination effect of the heating surface, and it predicts the experimental data with a mean relative error of 9.89%.

MODELING LOFW IN A PWR USING MELCOR

The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.