Experimental and Analytical Studies on Penetration Integrity of the Reactor Vessel Under External Vessel Cooling

Experimental and analytical studies on the penetration integrity of the reactor vessel in the APR (Advanced Power Reactor) 1400 have been performed under the condition of external vessel cooling in a severe accident. The objective of this study is to estimate failure or non-failure of the penetration including the ICI (In-Core Instrumentation) nozzle and the thimble tube. Five tests in conditions with and without external vessel cooling have been performed to estimate the effects of system, corium mass, and vessel geometry using alumina (Al2O3) melt as a simulant. The test results have been evaluated using the LILAC (Lower head IntegraL Analysis computer Code). The tests results have shown that penetration in the no external vessel cooling case is more damaged than that in the external vessel cooling case. An increase in system pressure from 0.9 MPa to 1.5 MPa was not effective on penetration damage, but an increase in corium mass from 40 kg to 60 kg and a vessel geometry change to flat plate with curvature were effective. The LILAC results are very similar to the test results on the ablation depth in the weld. It is concluded that external vessel cooling is a very effective means for maintaining penetration integrity.

Download Full-text

Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering ◽

10.1115/icone2020-16664 ◽

2020 ◽

Author(s):

R. Lo Frano ◽

S. Paci ◽

P. Darnowski ◽

P. Mazgaj

Keyword(s):

Nuclear Power ◽

Quantitative Estimation ◽

Computer Code ◽

Severe Accident ◽

Symmetric Model ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Term Operation ◽

Ageing Effects ◽

Lower Head

Abstract The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

Download Full-text

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

10th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone10-22176 ◽

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.

Download Full-text

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

Volume 3B: Design and Analysis ◽

10.1115/pvp2018-85101 ◽

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.

Download Full-text

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3D©

10th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone10-22754 ◽

2002 ◽

Cited By ~ 2

Author(s):

D. L. Knudson ◽

J. L. Rempe

Keyword(s):

Heat Transfer ◽

Molten Pool ◽

Reactor Vessel ◽

Severe Accident ◽

Complex Processes ◽

Lower Head ◽

Environmental Laboratory ◽

Molten Materials ◽

Core Materials

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D© has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D© relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D© are outlined.

Download Full-text

Development of the Penetration Tube Failure Analysis Program in the Lower Head of the Reactor Vessel During a Severe Accident

Nuclear Technology ◽

10.1080/00295450.2021.1929769 ◽

2021 ◽

pp. 1-16

Author(s):

J. Jung ◽

H. Y. Kim ◽

S. M. An

Keyword(s):

Failure Analysis ◽

Reactor Vessel ◽

Severe Accident ◽

Analysis Program ◽

Lower Head ◽

Tube Failure

Download Full-text

Research and Development of Validation and Drill System for Full Scope of Severe Accident Management Guideline

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4036892 ◽

2017 ◽

Vol 3 (4) ◽

Author(s):

Zhifei Yang ◽

Yali Chen ◽

Hu Luo

Keyword(s):

Research And Development ◽

Computer Code ◽

Severe Accident ◽

Accident Analysis ◽

Spent Fuel ◽

Accident Management ◽

Analysis Computer ◽

Spent Fuel Pool ◽

Accident Scenarios ◽

Drill System

To respond to the urgent needs of verification, training, and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities, and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under power condition, shutdown condition, spent fuel pool (SFP) condition, and refueling conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis computer code modular accident analysis program 5 (MAAP5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination, and verification.

Download Full-text

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

Volume 6: Beyond Design Basis Events; Student Paper Competition ◽

10.1115/icone21-15831 ◽

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.

Download Full-text

Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

Science and Technology of Nuclear Installations ◽

10.1155/2012/305405 ◽

2012 ◽

Vol 2012 ◽

pp. 1-8 ◽

Cited By ~ 5

Author(s):

Alejandro Nuñez-Carrera ◽

Raúl Camargo-Camargo ◽

Gilberto Espinosa-Paredes ◽

Adrián López-García

Keyword(s):

Pressure Vessel ◽

Reactor Core ◽

Cooling Water ◽

Reactor Vessel ◽

Reactor Pressure Vessel ◽

Severe Accident ◽

Core Material ◽

Detailed Model ◽

Lower Head ◽

Reactor Pressure

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

Download Full-text

Application of MELCOR to Westinghouse 4-loop PWR severe accident and evaluation of RPV lower head performance

Nuclear Science and Technology ◽

10.53747/jnst.v4i3.230 ◽

2014 ◽

Vol 4 (3) ◽

pp. 1-6

Author(s):

Dai Dien Le ◽

Thi Hoa Bui ◽

Thi Huong Vo

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Computer Code ◽

Severe Accident ◽

Ac Power ◽

Lower Head ◽

Active Safety Systems ◽

Station Blackout ◽

Control Volumes ◽

Safety Systems

In this study, MELCOR computer code is used to simulate the progression of a severe accident initiated from station blackout (SBO) accident for a Westinghouse 4-loop PWR. The hydraulic system is modeled using control volumes and flow paths. The reactor pressure vessel and internals, the primary loops with a pressurizer, steam generators, containment and accumulators are simulated for steady state in a good agreement with reference data. The two scenarios concerning SBO are investigated. The first scenario simulates RCP seal leakage during SBO and the other is SBLOCA to highlight an effectiveness of accumulators as well as to compare with the first simulation. All active safety systems which depend on AC power are assumed to be unavailable in this analysis. The main result of the study is an evaluation of RPV lower head integrity during severe accidents. This is preliminary work and expected to give the experience for further studies in the severe accident in nuclear power plants.

Download Full-text