Current perspectives on nuclear power plant risks and the use of risk-based information in regulation

For the past five years, the U.S. Nuclear Regulatory Commission has supported extensive studies of severe accidents. One outcome of this work is a set of advanced method for analyzing the probabilities, source terms, consequences, and risks of such accidents. These methods are being applied to a set of six U.S. commercial nuclear power plants, covering a wide spread of nuclear steam supply systems and containment designs. This work is to be documented in the Reactor Risk Reference Document, NUREG-1150, and supporting contractors reports. The methods being used for NUREG-1150, and some initial plant results, are briefly described in this paper.     

Impact research for containment design

Engaged for many years in research work concerning the safety and integrity of nuclear containments, the first author has performed numerous theoretical and experimental investigations at this institute. Airplane crashes on nuclear power plants, as well as containment attacks by detonation and missiles generated by bursting vessels have been studied with respect to practical design. Also, a series of fundamental researches has been done to evaluate constitutive laws for shockwaves in concrete and constitutive relations for concrete with regard to strain rate effects. Further investigations have focused on friction phenomena for projectiles impinging on concrete.     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

Evaluation of material flaws identified during fabrication leak testing of 304L containment vessel

Abstract

Savannah River National Laboratory has designed the bulk tritium shipping package as a replacement for the UC609 which was designed in the early 1970s. In the course of prototype package fabrication four components of the containment vessel boundary failed helium leak testing due to material flaws in the base material. Leak testing was being performed in accordance with ANSI N14·5, ‘Radioactive materials-leak tests on packages for shipment’. This paper addresses the ASME Section III test requirements for components and the failure mode of the base material. Additionally, the paper discusses the process used to correct the flaws for this application and recommendations for eliminating this potential in future applications.     

SNR 300 fast breeder reactor: steel containment - design, erection, testing

The Kalkar Nuclear Power Plant which is equipped with an 300 MW fast breeder reactor is being built by a Consortium mainly comprising German, Belgian and Dutch companies.The components of the fast breeder reactor are enclosed in a concrete containment which is designed to withstand severe external and internal loading.The concrete enclosure is surrounded by a steel containment which is designed to prevent the release of radioactivity following a postulated accident involving the nuclear components inside the concrete containment.The paper describes the solutions adopted for the different parts of the steel containment, the calculations verifying the suitability of the designs, the erection and the steel containment pressure and leak tests. The tests were performed with successful results in 1984.     

Experimental data base for containment thermalhydraulic analysis

This paper describes the joint research project DABASCO which is supported by the European Community under a cost-shared contract and participated by nine European institutions. The main objective of the project is to provide a generic experimental data base for the development of physical models and correlations for containment thermalhydraulic analysis. The project consists of seven separate-effects experimental programs which deal with new innovative conceptual features, e.g. passive decay heat removal and spray systems. The results of the various stages of the test programs will be assessed by industrial partners in relation to their applicability to reactor conditions.     

Evaluation of calculational and material models for concrete containment structures

A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measued strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given.     

Experiments and computational models for aerosol behaviour in the containment

Aerosols generated by condensation of volatile fission products during nuclear reactor core meltdown accidents represent a major fraction of the accidental airborne radioactivity. A comprehensive experimental research programme was performed at Battelle to investigate the transport and deposition behaviour of aerosols in the containment, in order to support the development of computer models which estimate the fission product behaviour in the containment and the source term for potential radionuclide releases to the environment. Important steps in the investigations were: (1) DEMONA experiments. The first large scale aerosol test series performed in the Battelle model containment (BMC) (total volume 640 m³), using an open (quasi one-room) geometry and condensation aerosols from a plasma torch generator. (2) VANAM experiments. Advanced aerosol tests in the BMC, using a multi-room geometry, mixed hygroscopic/non hygroscopic condensation aerosols, a double injection period, and varying thermohydraulic conditions. One of the experiments was subject of the International Standard Problem ISP 37. (3) KAEVER experiments. A systematic investigation of aerosol materials and mixtures and the related deposition behaviour, using a simplified one-room test vessel (10 m³ volume) and advanced instrumentation. Important computer codes developed and/or validated in connection with the experiments are FIPLOC and NAUA; aerosol codes CONTAIN, MELCOR and GOTHIC-MAEROS were also applied. Some important results from the investigations and code developments are: (1) significant local aerosol concentration differences can occur in a multi-room geometry; (2) concentration differences can be caused by atmospheric stratification; and (3) deposition is strongly affected by material hygroscopicity and atmospheric humidity. (4) Satisfactory prediction requires a consistent treatment of multi-room thermal hydraulics, aerosol transport and steam condensation on particles. (5) Prediction results can be affected by numerical stability and nodalization (user experience). This paper presents a number of results of the experimental investigations and the present state of code modelling, with special reference to the findings of ISP37.     

Concept study for a combined reinforced concrete containment

A variety of different types of steel and concrete containments have been designed and constructed in the past. Most of the concrete containments had been pre-stressed, offering the advantage of small displacements and a certain leak-tightness of the concrete itself. However, considerable stresses in concrete as well as in the tendons have to be maintained during the whole lifetime of the plant in order to guarantee the required pre-stressing. The long-time behaviour and the ductility in the case of beyond-design-load cases must be verified. Contrary to a pre-stressed containment a reinforced containment will only be significantly loaded during test conditions or when needed in case of an accident. It offers additional margins which can be used especially for dynamic loads such as impacts or for beyond-design events.The aim of this paper is to show the feasibility of a so-called combined containment which means a containment capable of resisting both severe internal accidents and external hazards, mainly the aircraft crash impact as considered in the design of nuclear power plants in Germany.The concept is based on a lined reinforced containment without pre-stressing. The mechanical resistance function is provided by the reinforced concrete and the leak-tightness function is provided by a so-called composite liner made of non-metallic materials. Some results of tests performed at Siemens laboratories and at the University of Karlsruhe which show the capability of a composite liner to bridge over cracks at the concrete surface will be presented in the paper.The study shows that the combined reinforced concrete containment with a composite liner offers a robust concept with high flexibility with respect to load requirements, beyond-design events and geometrical shaping (arrangement of openings, an integration of adjacent structures). The concept may be further optimized by partial pre-stressing at areas of high concentration of stresses such as at transition zones or at disturbances around large openings.     

Heat transfer from saturated and superheated atmospheres for containment analysis

Heat transfer correlations used in containment analysis are reviewed. Particular attention is paid to the Main Steam Line Break (MSLB) accident, which can lead to superheated atmospheres. It is shown that the standard condensating heat transfer correlations are inapplicable for such conditions. A new condensing-convective heat transfer model is derived on the basis of a mechanistic analysis of a vapour-steam boundary layer on a vertical condensing surface. The expression is normalized to experimental data in order to include the effect of forced convection. The new correlation is applicable to the superheated and saturated state and provides a smooth transition between both heat transfer regimes. It evaluates implicitly both condensing and convective heat transfer as a function of atmospheric and wall temperature conditions, and thus determines not only the total amount of heat transferred but also the amount of condensate formed.     

Research needs for design of concrete containment structures

Potential failure modes of reinforced concrete containment shells are outlined, especially those associated with pressure-induced cracking and seismic forces. A summary is given of experimental and analytical research needed to evaluate tangential shear capacity and stiffness, the interaction between liner and cracked concrete, peripheral (punching) shear capacity, radial shear behavior, and nonlinear dynamic analysis approaches.     

Treatment of hydrodynamic effects for toroidal containment vessels

Hydrodynamic effects in liquid-shell systems may be characterized in terms of structural degrees of freedom alone if an ideal fluid is assumed. The hydrodynamic effects are modeled by means of a consistent (full) added mass matrix which is obtained via finite element methods. The procedure is demonstrated for the case of a nuclear reactor toroidal containment vessel partially filled with water. Results demonstrate the superiority of this method over diagonal added mass procedures, such as the tributary area method.     

Creep prediction of concrete for reactor containment structures

Since the biggest time-dependent prestress loss of a prestressed concrete nuclear reactor containment structure is due to the creep of concrete, creep is one of the most important structural factors to be considered for the safety of a reactor containment structure during design, construction and maintenance. Creep in concrete has also recently been considered in evaluation of the crack resistance of concrete at an early-age in the durability examination of massive concrete structures like reactor containment structures. Existing empirical formulas on creep prediction show errors in their predictions due to simplified consideration of mixture proportions, and they also show large discrepancy among their predictions. In addition, they do not consider early-age behaviors of concrete and thus are mainly for the prediction of long-term creep at hardened concrete. In this paper, the creep characteristics of the reactor's both early-age and hardened reactor concrete made of type V cement are examined by carrying out both early-age and long-term creep tests. Then, the creep of the reactor concrete is predicted by using major creep-prediction equations of the AASHTO LRFD design specification, the Japanese standard specification for concrete structure, the ACI Committee 209 and the CEB/FIP model code and the Bazant and Panula's model, and the predicted results are compared with the test results. From the comparison, the applicability of the creep-prediction equations for the concrete of a reactor containment structure at both early-age and hardened stages is discussed.     

Methods for assessing NPP containment pressure boundary integrity

Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion.     

Numerical models for prestressing tendons in containment structures

Two modified stress–strain relations for bonded and unbonded internal tendons are proposed. The proposed relations can simulate the post-cracking behavior and tension stiffening effect in prestressed concrete containment structures. In the case of the bonded tendon, tensile forces between adjacent cracks are transmitted from a bonded tendon to concrete by bond forces. Therefore, the constitutive law of a bonded tendon stiffened by grout needs to be determined from the bond–slip relationship. On the other hand, a stress increase beyond the effective prestress in an unbonded tendon is not section-dependent but member-dependent. It means that the tendon stress unequivocally represents a uniform distribution along the length when the friction loss is excluded. Thus, using a strain reduction factor, the modified stress–strain curve of an unbonded tendon is derived by successive iterations. In advance, the prediction of cracking behavior and ultimate resisting capacity of prestressed concrete containment structures using the introduced numerical models are succeeded, and the need for the consideration of many influencing factors such as the tension stiffening effect, plastic hinge length and modification of stress–strain relation of tendon is emphasized. Finally, the developed numerical models are applied to prestressed concrete containment structures to verify the efficiency and applicability in simulating the structural behavior with bonded and/or unbonded tendons.     

Probability and containment of turbine missiles

With the trend toward ever larger nuclear power generating plants with large high-speed turbines, an important plant design consideration is the potential for and consequences of mechanical failure of turbine rotors. Such rotor failure could result in high-velocity disc fragments (turbine missiles) perforating the turbine casing and jeopardizing vital plant systems. The designer must first estimate the probability of any turbine missile damaging any safety-related plant component for his turbine and his plant arrangement. If the probability is not low enough to be acceptable to the regulatory agency, he must design a shield to contain the postulated turbine missiles. Alternatively, the shield could be designed to retard (to reduce the velocity of) the missiles such that they would not damage any vital plant system. In this paper, some of the presently available references that can be used to evaluate the probability, containment and retardation of turbine missiles are reviewed; various alternative methods are compared; and subjects for future research are recommended.     

Comparison between calculated and measured eigenfrequencies for spherical steel containment shells

Some eigenmodes and several hundreds of eigenfrequencies were measured for a very accurate spherical steel containment model. Even the influence of the surrounding air could be observed. Comparison with computational results shows surrounding air could be observed. Comparison with computational results shows shell response very well, provided the spatial resolution is sufficient. Based on these results the influence of shell imperfections which are typical for real containments will be investigated in the next step. Using the law of similarity all the results can easily be transferred from the model to the full scale containment.