Improvement of evaluation of the contact pressure upon the penetration of a reactor lower head during a severe accident: Part 1. Conceptualization considering elastic deformation

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.

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Numerical analysis of the transient wear and lubrication behaviors of misaligned journal bearings caused by linear shaft misalignment

Journal of Tribology ◽

10.1115/1.4051776 ◽

2021 ◽

pp. 1-17

Author(s):

Tianyou Yang ◽

Yanfeng Han ◽

Yijia Wang ◽

Guo Xiang

Keyword(s):

Numerical Model ◽

Film Thickness ◽

Contact Pressure ◽

Elastic Deformation ◽

Journal Bearing ◽

Fluid Pressure ◽

Axial Direction ◽

Wear Depth ◽

Wear Process ◽

Journal Misalignment

Abstract The purpose of this study is to investigate the role of the misalignment journal, caused by journal elastic deformation, on the transient wear and mixed lubrication performances using a numerical model. In the numerical model, the transient geometry lubrication clearance considering the journal misalignment, the transient elastic deformation and the transient wear depth are incorporated to evaluate the transient film thickness during wear process. The evolutions, under different external loads, of the wear depth, wear rate, elastic deformation, film thickness, fluid pressure and contact pressure are calculated by the numerical model. Furthermore, the calculated results of the misaligned journal bearing are compared with those of the aligned journal bearing. The results show that the distributions of the wear depth, film pressure and elastic deformation are asymmetric along the axial direction and the peak values of them shift toward the back end when the journal misalignment is considered. The maximum wear depth, maximum fluid pressure, maximum contact pressure and maximum elastic deformation of the misaligned journal condition are significantly larger than those of the aligned journal condition.

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Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2004.11.013 ◽

2005 ◽

Vol 235 (8) ◽

pp. 845-853 ◽

Cited By ~ 7

Author(s):

V. Koundy ◽

I. Cormeau

Keyword(s):

Analytical Modeling ◽

Shell Theory ◽

Severe Accident ◽

Lower Head ◽

Accident Conditions

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Results of the QUENCH-L2, DISCO-L2, and COMET-L2 Experiments Performed Within the LACOMERA Project at the Forschungszentrum Karlsruhe

Volume 5: Safety and Security; Low Level Waste Management, Decontamination and Decommissioning; Nuclear Industry Forum ◽

10.1115/icone14-89343 ◽

2006 ◽

Author(s):

Alexei Miassoedov ◽

Hans Alsmeyer ◽

Leonhard Meyer ◽

Martin Steinbrueck ◽

Pavlin Groudev ◽

...

Keyword(s):

Large Scale ◽

Fluid Dynamic ◽

Test Phase ◽

Severe Accident ◽

Lower Head ◽

Dry Conditions ◽

Experimental Facilities ◽

Reactor Types ◽

The Eu

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: Fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.

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Ex-Vessel Coolability Analysis for a Belgian NPP

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications ◽

10.1115/icone17-75456 ◽

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.

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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.

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Expansion of high temperature creep test data for failure evaluation of BWR lower head in severe accident

Mechanical Engineering Journal ◽

10.1299/mej.19-00560 ◽

2020 ◽

Vol 7 (3) ◽

pp. 19-00560-19-00560

Author(s):

Yoshihito YAMAGUCHI ◽

Jinya KATSUYAMA ◽

Yoshiyuki KAJI ◽

Masahiko OSAKA ◽

Yinsheng LI

Keyword(s):

High Temperature ◽

Test Data ◽

Creep Test ◽

Severe Accident ◽

High Temperature Creep ◽

Temperature Creep ◽

Lower Head ◽

Failure Evaluation

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Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments

Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy ◽

10.1115/icone14-89159 ◽

2006 ◽

Author(s):

Larry L. Humphries ◽

Tze Yao Chu ◽

John H. Bentz

Keyword(s):

Wall Temperature ◽

Pressure Vessel ◽

Severe Accident ◽

Core Material ◽

Material Strength ◽

Test Design ◽

Temperature Differential ◽

Pressurized Water ◽

Accident Management ◽

Lower Head

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW, is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.

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