Severe Accident Loads Research Articles

Effects of steel liner corrosion on the leak-tightness of prestressed concrete containment structure under severe accident loads

This study evaluates the effect of steel liner corrosion on the leak-tightness of the prestressed concrete containment structure under severe accidents. Based on the nonlinear finite element method and sub-model technique in the ABAQUS platform, a global model of the containment structure and a detailed sub-model of the corroded steel liner are established. Different calculation methods for the multiaxial stress factor in the failure criterion of the steel liner are compared, and the failure criterion is modified by combining the uniaxial tensile test results of corroded steel liners. According to the modified failure criteria, the corrosion shape and corrosion degree of the steel liner, as well as the thermal effect on the containment structure, are studied in depth. Results indicate that the mechanical properties of the steel liner degrade significantly after corrosion, and the simulation method proposed in this study can reflect the corrosion state of the steel liner well. The multiaxial stress factor based on the Davis triaxiality coefficient is conservative in evaluating functional failure of the containment structure, and the calculation method of the corrosion factor proposed in this study can directly reflect the effect of the corrosion degree on the ductility of the steel liner. Considering that a less conservative calculation result will be obtained based on regular corrosion shapes, the corrosion shapes established from actual corrosion cases of steel liners in this study are available for reference. Pressure capacity of the containment structure corresponding to functional failure decreases with the increase in corrosion degree of the steel liner. When corrosion rates range from 10% to 50%, pressure capacity under the only pressure case and the coupled temperature and pressure case is decreased by 9.21% to 28.58% and 11.35% to 31.06%, respectively. The behavior of the steel liner is significantly affected by the temperature under corrosion conditions.

Effects of Steel Liner Corrosion on the Leak-Tightness of Prestressed Concrete Containment Structure Under Severe Accident Loads

Effects of Steel Liner Corrosion on the Leak-Tightness of Prestressed Concrete Containment Structure Under Severe Accident Loads

Nonlinear Finite Element Analysis of Prestressed Concrete Containment Vessel under Severe Accident Loads

Prestressed concrete containment vessel act as a reliable leak tight barrier during the accident conditions. There is a growing demand to study nonlinear behavior of containment structure under severe accident loads in depth. This paper presents nonlinear finite element analysis of prsestressed concrete containment vessel under severe accident loads with consideration of material nonlinearity, penetrations, local reinforcement and temperature-dependent degradation characteristic of materials. To reflect the prestressing effects realistically, non-uniform distribution of effective prestressing along the tendon profile is explicitly considered and python scripts are developed to add the corresponding temperature drop value for each node of the prestressed tendons to ABAQUS input files automatically. Nonlinear finite element analysis for pressure only case and combined thermal and pressure case has been investigated in detail. Nonlinear finite element analysis results of the containment structure indicate that, thermal effects have negligible effect on pressure capacity of containment, considering the thermal effects, pressure capacity of containment decrease not more than 5% and margin of the containment still meets the requirements of not less than 2.5. The effect of temperature exhibit significantly influence on displacement response of containment structure, and the effect of temperature on the strain of liner and reinforcing steel is much greater than that on prestressed tendon. Thermal effects exhibit the greatest influence on nonlinear displacement response of the dome apex location and the least influence on the 33 m elevation.

Development of radiation risk assessment simulator using system dynamics methodology

The potential magnitudes of radionuclide releases under severe accident loadings and offsite consequences as well as the overall risk (the product of accident frequencies and consequences) are analyzed and evaluated quantitatively in this study. The system dynamics methodology has been applied to predict the time-dependent behaviors such as feedback and dependency as well as to model uncertain behavior of complex physical system. It is used to construct the transfer mechanisms of time dependent radioactivity concentration and to evaluate them. Dynamic variations of radio activities are simulated by considering several effects such as deposition, weathering, washout, re-suspension, root uptake, translocation, leaching, senescence, intake, and excretion of soil. The time-dependent radio-ecological model applicable to Korean specific environment has been developed in order to assess the radiological consequences following the short-term deposition of radio-nuclides during severe accidents nuclear power plant. An ingestion food chain model can estimate time dependent radioactivity concentrations in foodstuffs. And it is also shown that the system dynamics approach is useful for analyzing the phenomenon of the complex system as well as the behavior of structure values with respect to time. The output of this model (Bq ingested per Bq m−2 deposited) may be multiplied by the deposition and a dose conversion factor (Gy Bq−1) to yield organ-specific doses. The model may be run deterministically to yield a single estimate or stochastic distributions by “Monte-Carlo” calculation that reflects uncertainty of parameter and model uncertainties. The results of this study may contribute to identifying the relative importance of various parameters occurred in consequence analysis, as well as to assessing risk reduction effects in accident management.

Study of tearing behaviour of a PWR reactor pressure vessel lower head under severe accident loadings

In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident. Both the LHF and OLHF experiments as well as the FOREVER experiments revealed that initiation of the failure is typically local. For the case of a uniform temperature distribution in the lower head, crack initiation occurs in the thinnest region and for the case of a non-uniform temperature distribution, it initiates at the highest temperature region. These experimental results can be modelled numerically (but more accurately with 3D finite element codes). The failure time predictions obtained using numerical modelling agree reasonably well with the experimental values. However, the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results). Indeed, the size of breach depends on the mode of crack propagation which is directly related to the metallurgical characteristics of the RPV steel. Small changes in the initial chemical composition of the vessel material can lead to different types of rupture behaviour at high temperatures. Different rupture behaviours were observed in the LHF and OLHF experiments using the SA533B1 steel. Similar observations were previously noticed during a CEA material characterization programme on the 16MND5 steel. To determine crack propagation and final failure size, 3D modelling would thus be needed with an adequate failure criterion taking into account the variability in behaviour of the RPV material at high temperatures. This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA Lyon. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions. This programme was initiated in 2003 and is made up of five main sections, namely an inventory of the different French PWR lower head materials, metallurgical investigations to better understand the cause of mechanical behaviour variability that is observed and related to material microstructure, Compact Tension (CT) testing of specimens to characterize the tear resistance of the material, validation of the modelling using experiments on tube specimens and the development of a new failure criterion for the 3D finite element models.

On the prediction of the reactor vessel integrity under severe accident loadings (RPVSA)

In order to allow more reliable predictions on the lower head response under core melt-down conditions, the temperature distribution has been analysed including the natural convection in the corium pool. Furthermore, the mechanical models and the failure criteria have been improved based on the RUPTHER and FASTHER experiments where typical temperature gradients are simulated. Lower head local melting as well as corium crust development has been addressed in the CORVIS experiments studying the contact between an alumina/iron thermite and a thick steel plate. The upper head loading by corium impact due to a postulated in-vessel steam explosion has been investigated by the BERDA experiments. Similarity rules were considered such that the results can be directly converted to reactor conditions. Based on these investigations admissible steam explosion energy releases are determined which the upper head can carry. If these limits are not exceeded the reactor containment cannot be endangered by broken head fragments. To provide the necessary basic data, mechanical material tests have been performed.

Insights and comparisons of the level-2 results of recent probabilistic safety analyses

Observations and insights based on the review of a number of recent level-2 probabilistic safety analyses (PSAs) and individual plant examinations (IPEs) are provided. Observations and comparisons are made regarding plant and containment design characteristics, methods for the analysis of containment response to severe accident loads, modeling of the uncertain phenomenological processes impacting containment response, accident progression and containment analysis, source term calculation, and uncertainty analysis. Insights are obtained which attempt to relate the various plant and containment design characteristics to expected containment performance, though these relationships can often be obscured by the large inherent uncertainties associated with quantification of most level-2 PSA issues.

Current state of knowledge on the behavior of steel liners in concrete containments subjected to overpressurization loads

In the US, concrete containment buildings for commercial nuclear power plants have steel liners that act as the internal pressure boundary. The liner abuts the concrete, acting as the interior concrete form. The liner is attached to the concrete by either studs or by a continuous structural shape (such as a T-section or channel) that is either continuously or intermittently welded to the liner. Studs are commonly used in reinforced concrete containments, while prestressed containments utilize a structural element as the anchorage. The practice in some countries follows the US practice, while in other countries the containment does not have a steel liner. In this latter case, there is a true double containment, and the annular region between the two containments is vented. This paper will review the practice of design of the liner system prior to the consideration of severe accident loads (overpressurization loads beyond the design conditions). An overpressurization test of a 1:6 scale reinforced concrete containment at Sandia National Laboratories resulted in a failure mechanism in the liner that was not fully anticipated. Post-test analyses and experiments have been conducted to understand the failure better. This work and the activities that followed the test are reviewed. Areas in which additional research should be conducted are given.

Design considerations for concrete containments under severe accident loads

Many existing containments in the United States have been shown to accommodate credible severe accident loads. Future containments should be explicitly designed for severe accident loads to reduce the uncertainty associated with the response of containments to these low-probability events. This paper examines the experiences from the application of current structural design codes for concrete containments, ultimate pressure capacity evaluation of existing containments, and pressure fragility testing of scale model concrete containments to arrive at the directions for modification of national codes. Recommendations are provided to consider the severe accidents directly in the concrete containment design.

A generic approach for steel containment vessel success criteria for severe accident loads

Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. Department of Energy (DOE), Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. This paper is limited to a discussion of a generic approach for steel containment vessel success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the “best estimate” of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks.

Research and Development to Improve Containment for the Next Generation of Pressurised Water Reactor Plants

In recent years politicians and scientists have increasingly demanded that, on account of the serious consequences resulting from a core meltdown accident, future nuclear power plants should include additional safety features. This requirement has been taken up by the industry and improved containment concepts for future pressurised water reactors (PWRs) are being considered and appropriate R&D studies on severe accident loadings conducted. These include the estimation of loads which might result from an energetic in vessel steam explosion, high pressure failure of the reactor pressure vessel, or dynamic hydrogen combustions. Conservative estimates of the upper limits of loading are presented. Complementary investigations concentrate on reliable, long term removal of the decay heat from the core melt after an accident, and from the accident atmosphere in the containment. In this context, two core melt cooling concepts (core catchers) are presented which can serve as innovative elements in future PWR plants. The general goal of these studies consists in contributions to be made to the development of a containment which will not allow significant radiological consequences for the environment, even under the extreme conditions of severe reactor accidents.

Performance of containment penetrations under severe accident loadings

The paper provides a summary of efforts to date to better understand the leakage behavior of containment penetrations when subjected to severe accident conditions. The research activities discussed herein are a part of the Containment Integrity Programs, which are managed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Past containment penetration research topics, which are briefly described, include testing of typical compression seals and gaskets, electrical penetration assemblies, and a personnel airlock, as well as an investigation of leakage due to ovalization of penetration sleeves. The primary focus of the paper is on recent or ongoing research programs on the behavior of inflatable seals, bellows, and of pressure unseating equipment hatches.

Construction of a [formula omitted] concrete containment model

This paper discusses the features and construction of a 1 6 - scale reinforced-concrete containment model that has been built at Sandia National Laboratories in Albuquerque, New Mexico. The model Light-Water-Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc. The containment model will be tested to failure to determine its response to static internal overpressurization. The results from testing the heavily instrumented containment will be used to assess the capability of analytical methods for predicting the performance of containments subject to severe accident loads as part of the US Nuclear Regulatory Commission's program on containment integrity. The scaled dimensions of the cylindrical wall and hemispherical dome are typical of a full-size containment. Features representative of a prototypical containment and included in the heavily reinforced model are equipment hatches, personnel airlocks, several small piping penetrations, and a thin steel liner attached to the concrete by headed studs.

Failure mechanisms of LWR steel containment buildings subject to severe accident loadings

The potential failure mechanisms in LWR steel containment buildings subject to quasi-static pressurization and elevated temperature are identified. For each failure mechanism, the relevant structural response measures are discussed. For mechanisms involving leakage, the importance of seal performance is also discussed. Criteria that can be used to evaluate threshold environments are presented for several failure mechanisms. Results of tests on scale models and seal tests that support the criteria are referenced.