Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

Structural Integrity of a Reactor Vessel Lower Head Under In-Vessel Steam Explosion Loads

2013 ◽  
Author(s):  
Guohong Xue ◽  
Yinbiao He ◽  
Ming Cao ◽  
Hao Yu ◽  
Yongjian Gao
Keyword(s):  
Steam Explosion ◽  
Power Plants ◽  
Structural Integrity ◽  
Equivalent Plastic Strain ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Conservation Of Energy ◽  
Lower Head ◽  
Pressure Pulses

Passive nuclear power plants emphasize the “In vessel retention” idea such that, after a postulated severe accident event, the reactor vessel wall, flooded with emergency cooling water, will maintain its structural integrity and consequently keep the molten core inside the reactor vessel. However, steam explosion may still occur when the melting core or molten metal is mixed with cooling water. The huge pressure pulses from the steam explosion may be a threat to the structural integrity of the reactor vessel lower head and the potential failure may make the situation difficult to control. This paper presents a detailed analysis on the structural integrity of a reactor vessel lower head. First, a mathematical model is built to relate the equivalent plastic strain in the lower head under explosive loads based on the law of conservation of energy. Then a finite element model, using the computer code ABAQUS, is built and the material’s yield strength as a function of strain rate was simulated using the Bodner-Symonds methodology. With this model, the dynamic response and the structural integrity of the reactor vessel lower head is studied, considering the effect of the magnitude, the shape and the duration of the pressure pulses. The method used in this paper is believed to be applicable to other types of devices containing potential explosive materials and thus could provide guiding significance to similar problems.

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

Materials Aging Degradation of Reactor Vessel Internals: Part III—Structural Evaluation and Local Finite Element Models

2009 ◽  
Author(s):  
Matthew D. Snyder ◽  
Tama´s R. Liszkai ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Structural Analysis ◽  
Constitutive Model ◽  
Reactor Vessel ◽  
Finite Element Models ◽  
Local Models ◽  
Bolted Connections ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) internals components can experience material aging and degradation due to irradiation. The purpose of the functionality analysis is to provide a best-estimate evaluation of the reactor internals core barrel assembly for materials degradation to see if the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis [1] representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox-type (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates of the internals. The computational fluid dynamics domain (CFD) allows evaluation of the internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections. Therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [2] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, to be presented at PVP 2009 [2] describes global structural finite element models. Part III, presented in this paper, presents a description of local models of bolted connections, results, and conclusions.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

Ex-Vessel Coolability Analysis for a Belgian NPP

2009 ◽  
Author(s):  
Jarne R. Verpoorten ◽  
Miche`le Auglaire ◽  
Frank Bertels
Keyword(s):  
State Of The Art ◽  
The State ◽  
Severe Accident ◽  
Research Projects ◽  
Specific Data ◽  
The Core ◽  
Core Damage ◽  
Accident Management ◽  
Lower Head ◽  
Core Cooling

During a hypothetical Severe Accident (SA), core damage is to be expected due to insufficient core cooling. If the lack of core cooling persists, the degradation of the core can continue and could lead to the presence of corium in the lower plenum. There, the thermo-mechanical attack of the lower head by the corium could eventually lead to vessel failure and corium release to the reactor cavity pit. In this paper, it is described how the international state-of-the-art knowledge has been applied in combination with plant-specific data in order to obtain a custom Severe Accident Management (SAM) approach and hardware adaptations for existing NPPs. Also the interest of Tractebel Engineering in future SA research projects related to this topic will be addressed from the viewpoint of keeping the analysis up-to-date with the state-of-the art knowledge.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC

2009 ◽  
Author(s):  
Polina Tusheva ◽  
Nils Reinke ◽  
Eberhard Altstadt ◽  
Frank Schaefer ◽  
Frank-Peter Weiss ◽  
...  
Keyword(s):  
Power System ◽  
Failure Time ◽  
Late Phase ◽  
Electric Power System ◽  
Physical Models ◽  
Severe Accident ◽  
Feed Water ◽  
The Core ◽  
Lower Head ◽  
Station Blackout

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

Sign in / Sign up

Export Citation Format

Share Document