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.     

Investigation of large-scale gas mixing and stratification phenomena related to LWR containment studies in the PANDA facility

The thermal-hydraulic processes governing the containment response to postulated accidents and the mixing and distribution of hydrogen following a severe accident are relevant issues identified by several international expert groups as ‘research needs’ for current and advanced LWRs. The development and validation of modern computational codes that will accurately predict gas distribution in LWR containments is required for analyses related to safety, design and operational issues in current and advanced reactors. This objective requires the availability of separate-effect test data collected in facilities where the 3D distribution of the relevant variables is measured with sufficient resolution and accuracy and tests are performed under well-controlled initial and boundary conditions. Within the scope of the OECD project ‘SETH’, a series of 25 tests has been initiated in the large-scale thermal-hydraulic facility PANDA, in order to investigate mixing and stratification phenomena in a large multi-compartment gas volume approaching the dimensions of actual containment compartments. This experimental programme investigates jets and plumes and the resulting propagation of stratification fronts. The presentation of the first results from this test series demonstrates the value of these new data for code validation purposes.     

Pretest predictions for the response of a 1:8 scale steel LWR containment building model to static overpressurization

The analyses used to predict the behavior of a 1:8-scale model of a steel LWR containment building to static overpressurization are described and results are presented. Finite strain, large displacement, and nonlinear material properties were accounted for using finite element methods. Three-dimensional models were needed to analyze the penetrations, which included operable equipment hatches, personnel lock representations, and a constrained pipe. It was concluded that the scale model would fail due to leakage caused by large deformations of the equipment hatch sleeves.     

Research related to buckling design of nuclear containment

This paper discusses the recent experimental and analytical studies related to buckling design of fabricated steel shells. The effects of initial imperfections and residual stresses on buckling are under investigation. The test programs include ring and stringer stiffened as well as ring stiffened cylinders subject to combinations of axial compression and external pressure. Proposed modifications to ASME Code Case N-284, “Metal Containment Shell Buckling Design Methods,” as well as the need for additional research, are discussed.     

Refined cracked concrete analysis of concrete containment structures subject to operational and environmental loadings

Recent commercial nuclear power plant containment concepts involve the use of large reinforced concrete structures to form pressure boundaries. Where these structures are not provided with an integral steel liner, excessive cracking of the concrete under loads could result in the loss of the pressure boundary integrity with the risk of over-pressurization of other structures. Cracking of concrete is a local phenomenon and considerable detail must be included in any analytical model to obtain sufficiently refined results for the prediction of crack size and propagation. This imposes severe limitations on the overall size of structures or structural components for which detailed cracking analysis can be considered directly. To overcome this restriction, a two step procedure was developed in which linear analyses were performed to obtain the gross response, and nonlinear cracking analyses were performed for selected portions of the structure to evaluate local cracking in detail. Through iteration, compatibility of behavior between the linear and nonlinear analyses was achieved with the gross response being used to extrapolate the local cracking results to predict cracking over the entire structure. This paper discusses the analysis procedures for the detailed evaluation of cracking in large reinforced concrete structures and components. Analyses performed for an actual unlined reinforced concrete containment structure using these procedures are discussed and results are presented.     

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.     

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.     

Limit analysis and design of containment vessels

In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness.In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon.The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed.A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods.The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the “limit states” design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper.     

Sliding connections for separating different pressures and temperatures in the HTR

Sliding connections can be used at such positions where gas compartments with different pressures and temperatures must be seperated from each other. These connections should be able to compensate different temperatures during operation and to accommodate axial and angular offsets and vibrations with connecting pipes. Determination of the leakage behaviour and the reliability of the system necessitates design work, computations and experimentation. The functional reliability and approval qualification capability of the system have already been demonstrated with the THTR. The permissible leakage values can be guaranteed for every diameter, based on tested sliding connections up to 3000 mm diameter. The design-determinant influencing parameters have been located.     

Stress corrosion cracking of Zircaloy-4 cladding at elevated temperatures and its relevance to transient LWR fuel rod behaviour

In extensive out-of-pile experiments from 500 to 900° C it has been shown that, of all the volatile fission products in a LWR fuel rod, only iodine can cause low ductility failure of Zircaloy-4 tubing due to stress corrosion cracking up to about 800° C. The critical iodine concentration above which brittle cladding failure occurs was determined as a function of temperature in the absence and presence of UO₂ fuel. A comparison of these values with the amount expected in the fuel cladding gap during a LOCA transient shows that a clear influence of iodine on burst strain can be expected only up to 700° C. This is in agreement with the results of in-pile LOCA tests performed in the FR-2 reactor with high burnup fuel rods. Since the burst temperatures during a LOCA transient would generally be above 700° C, an influence of iodine on burst strain is not very probable in a LOCA. However, with respect to ATWS transients where the maximum cladding temperatures would be below 700° C, an influence of iodine on the mechanical properties of zircaloy can be expected.     

Research requirements for improved design of reinforced concrete containment structures

Reinforced concrete is a competitive material for the construction of nuclear power plant containment structures. However, the designer is constrained by limited data on the behavior of certain construction details which require him to use what may be excessive rebar quantities and lead to difficult and costly construction. This paper discusses several design situations where research is recommended to increase the designer's options, to facilitate construction, and to extend the applicability of reinforced concrete to such changing containment requirements as may be imposed by an evolving nuclear technology.     

Serviceability design load factors and reliability assessments for reinforced concrete containment structures

A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments.     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

The electrochemistry of zirconium in aqueous solutions at elevated temperatures and pressures

The electrochemistry of zirconium has been explored in borate buffer solution of pH = 6.94 at 250 °C with and without hydrogen by measuring the current, impedance, and capacitance as a function of potential. Data are interpreted in terms of modified point defect models (PDM) that recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH₂ is predicted to be spontaneous rather than the ZrO₂. Mott–Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer.     

Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

Detailed representations of the reactor core generate computational meshes with a high number of cells where the fluid dynamics equations must be solved. An exhaustive analysis of the CPU times needed by the thermal-hydraulic subchannel code COBRA-TF for different stages in the solution process has revealed that the solution of the linear system of pressure equations is the most time consuming process. To improve code efficiency two optimized matrix solvers, Super LU library and Krylov non-stationary iterative methods have been implemented in the code and their performance has been tested using a suite of five test cases. The results of performed comparative analyses have demonstrated that for large cases, the implementation of the Bi-Conjugate Gradient Stabilized (Bi-CGSTAB) Krylov method combined with the incomplete LU factorization with dual truncation strategy (ILUT) pre-conditioner reduced the time used by the code for the solution of the pressure matrix by a factor of 20. Both new solvers converge smoothly regardless of the nature of simulated cases and the mesh structures and improve the stability and accuracy of results compared to the classic Gauss–Seidel iterative method. The obtained results indicate that the direct inversion method is the best option for small cases.     

Results of a long-term wet storage demonstration test with intact and operational defective LWR fuel rods

28 spent fuel rods — 18 intact and 10 operational defective rods — were included in the storage test program. Within 7 years the spent fuel rods were inspected four times. To characterize the spent fuel rods the following methods were applied during pool inspections: visual inspection, profilometry, eddy current testing, and oxide thickness recording.Summarizing the results of the intermediate and of the final inspection it has to be concluded that — as predicted — no change exceeding the detection limit could be found either at the intact or at the operational defective fuel rods. These results must be regarded as conservative because handling of the different spent fuel rods during inspection provided additional and atypical loads — especially for the operational defective spent fuel — in comparison with the long term storage of complete fuel bundles.The results of this carefully documented demonstration test has shown agreement with the theoretical analysis and with the overall experience available from pool storage that wet spent LWR-fuel storage can be performed without any problems even for extended periods of time.     

The design and evaluation of a passively cooled containment for a high-rating pressurized water reactor

An integrated pressurized water reactor (PWR) containment was conceptualized that allows heat to be rejected passively to the environment. The proposed containment is based on the demonstrated Ebasco Waterford 3 design. The secondary concrete shell was equipped with inlet and outlet vents that create an air-convection annulus. These vents also permit the submersion of the lower part of the primary containment into an external water pool. An internal water pool located at the bottom of the lower containment was added to increase in-containment heat storage. The performance of the proposed passively cooled containment was evaluated using a subdivided volume code, version 3.4e; the relative novelty of subdivided volume analyses for containment performance evaluation requires experimental verification of principal code predictions. Two experiments were carried out; one to test the performance of the external moat, and one to verify the code’s ability to predict thermal-stratification inside the containment. To improve the subdivided-volume simulation of convection-related parameters, a modeling technique (boundary layer flow approximation) was devised. Finally, the behavior of the proposed containment was evaluated for the worst-case large break loss of coolant accident and the worst-case main steam line break accident. Peak pressures remained below 0.45 MPa during both transients; internal wall pressure differences, equipment qualification temperatures, pressure restoration time also remained below design limits. The mitigation capability of hydrogen recombiners was also evaluated.     

Expansion of the method of correlated temperatures to clad and fuel design analyses

In the thermal design of nuclear reactor cores, specified design limits (temperatures and linear power rating) should not be exceeded by the operating values of certain elements (coolant, clad and fuel). However, a certain number of channels or fuel pins could be permitted to exceed the specified limits without affecting the reactor's safety while still allowing reliable operation. An expansion of the method of correlated temperatures, developed for coolant temperature analysis, was performed to enable clad temperature and fuel centerline melting analyses for reactor core reliability studies. Since generation of random numbers is involved, calculational procedures, tailored to designer needs, were developed in order to reduce computational time. The method is applied to a typical LMFBR core and results are presented for various assumed clad and fuel design limits.