Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianfeng Mao ◽  
Jianwei Zhu ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Heat Flux ◽  
Internal Pressure ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Wide Range ◽  
Depth Analysis ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

Enhancement of RPV Critical Heat Flux by Injection Flow

2017 ◽  
Author(s):  
Wei Chen ◽  
Canhui Sun ◽  
Jun Geng
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Metallic Layer ◽  
Fluid Temperature ◽  
Injection Flow ◽  
Severe Accidents ◽  
Reactor Pressure ◽  
Residual Heat

Under severe accidents, the reactor pressure vessel is flooded with water and the residual heat is removed by two-phase natural circulation through the flow channel between the reactor vessel and thermal insulation. If the heat flux of the outer wall is lower than local critical heat flux, the residual heat can be removed, and if the heat flux of the outer wall is higher than local critical heat flux, the reactor pressure vessel should be molten. For AP-type reactors, like AP1000 and CAP1400, critical heat flux of the reactor pressure vessel is the heat transfer limits of residual heat under severe accidents. Previous studies indicate that after severe accident a two-layer molten pool can be formed, namely metallic layer and oxide layer. Compared with oxide layer, in metallic layer, the heat flux more easily exceeds the heat transfer limits due to its low thermal resistance. In this study, an approach was proposed to enhance local critical heat flux. This approach is expected to be used in local area around reactor pressure wall, like metallic layer, to increase the reactor pressure vessel intact probability under severe accidents. In this new approach, injection flow channels are added to the present flow channel by adding simple flow pipes from insulation near 60 to 80 degree where exceeding critical heat flux is most likely to happen. The fluid flow under external reactor vessel cooling (ERVC) condition is divided into two parts: one part is from downward (0 degree) to upward (90 degree) along the curved reactor pressure vessel and the other part is from injection pipe (about 70 degree) to upward. The fluid temperature from injection pipe is lower than that from downward due to residual heat from the reactor pressure wall. And hence, the local critical heat flux is likely to increase because of inject turbulence and low fluid temperature. An experimental facility is conducted to study the mechanism of injection influence on critical heat flux under normal pressure condition. There are two main loops in this facility: one is main loop while the other is injection loop. The test section is an inclined downward heated rectangular channel with its inclined angle varied from 0 degree to 90 degree. Flow and thermal conditions are listed: in main loop, mass flow velocity ranging from 100kg/m2s to 600kg/m2s with fluid temperature from 90 °C to 105 °C; In injection loop, mass flow velocity ranging from 0 to 600 kg/m2s with fluid temperature from 85 °C to 105 °C. Under the above condition, with and without injection flow, critical heat flux experiments were conducted. It indicates that injection velocity has great effect on critical heat flux, while injection subcooled has little effect. The critical heat flux can be increased by 0.07MW/m2 to 0.33MW/m2 depending on various injection velocities and main loop conditions.

Comparative Study on Reactor Pressure Vessel Failure Behaviors With Various Geometric Discontinuities Under Severe Accident

2017 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianwei Zhu ◽  
Jianfeng Mao ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Comparative Study ◽  
Pressure Vessel ◽  
External Surface ◽  
Safety Margin ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Wide Range ◽  
Core Meltdown ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a basic strategy for severe accident (SA) mitigation of some advanced nuclear power plants (NPPs). The IVR strategy is to keep the reactor pressure vessel (RPV) intact under SA like core meltdown condition. During the IVR, the core melt (∼1327 °C) is collected in the lower head (LH) of the RPV, while the external surface of RPV is submerged in the water. Through external cooling of the RPV, the structural integrity is assumed to be maintained within a prescribed period of time. The maximum thermal loading is referred to critical heat flux (CHF) on the inside, while the external surface is considered to perform in the environment of the boiling crisis point (∼130 °C). Due to the high temperature gradients, the failure mechanisms of the RPV is found to span a wide range of structural behaviors across the wall thickness, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Besides CHF, the pressurized core meltdown was another evident threat to the RPV integrity, as indicated in the Fukushima accident on 2011. In illustrating the effects of internal pressures and individual CHF on the failure behaviors, three typical RPVs with geometric discontinuity caused by local material melting were adopted for the comparative study. Through finite-element method (FEM), the RPV structural behaviors were investigated in terms of deformation, stress, plastic strain, creep, and damage. Finally, some important conclusions are summarized in the concluding remark. Such comparative study provides insight and better understanding for the RPV safety margin under the IVR condition.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

Manufacturing and Properties of Closure Head Forging Integrated With Flange for PWR Reactor Pressure Vessel

2004 ◽  
Author(s):  
Komei Suzuki ◽  
Etsuo Murai ◽  
Yasuhiko Tanaka ◽  
Iku Kurihara ◽  
Tomoharu Sasaki ◽  
...  
Keyword(s):  
Weld Joint ◽  
Pressure Vessel ◽  
Material Properties ◽  
Impact Test ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Conventional Design ◽  
Manufacturing Technologies ◽  
Reactor Pressure

Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange.

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

2008 ◽  
Author(s):  
Kiminobu Hojo ◽  
Naoki Ogawa ◽  
Yoichi Iwamoto ◽  
Kazutoshi Ohoto ◽  
Seiji Asada ◽  
...  
Keyword(s):  
Residual Stress ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Stress Field ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Complex Structure ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

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