Tension tests of concrete containment wall elements

Tension tests of concrete containment wall elements were conducted as part of a three-phase research program sponsored by the Electric Power Research Institute (EPRI). The objective of the EPRI experimental/analytical program is twofold. The first objective is to provide the utility industry with a test-verified analytical method for making realistic estimates of actual capacities of reinforced and prestressed concrete containments under internal over-pressurization from postulated degraded core accidents. The second objective is to determine qualitative and quantitative leak rate characteristics of typical containment cross-sections with and without penetrations. This paper covers the experimental portion the the EPRI program.The testing program for Phase 1 included eight large-scale specimens representing elements from the wall of a containment. Each specimen was 60-in (1525-mm) square, 24-in (610-mm) thick, and had full-size reinforcing bars. Six specimens were representative of prototypical reinforced concrete containment designs. The remaining two specimens represented prototypical prestressed containment designs.Various reinforcement configurations and loading arrangements resulted in data that permit comparisons of the effects of controlled variables on cracking and subsequent concrete/reinforcement/liner interaction in containment elements.Subtle differences, due to variations in reinforcement patterns and load applications among the eight specimens, are being used to benchmark the codes being developed in the analytical portion of the EPRI program.Phases 2 and 3 of the test program will examine leak rate characteristics and failure mechanisms at penetrations and structural discontinuities.     

Tangential shear tests of reinforced concrete containment elements

An experimental program to investigate the behavior of large scale reinforced concrete elements subjected to biaxial tension and shear forces is described. Six tests are being conducted on specimens 5 ft(1.52 m) square and 2 ft(0.61 m) thick, with no. 14 and no. 18 reinforcement. Program variables are level of biaxial tension, and monotonic vs. reversing shear load. Two monotonic tests have been completed to date. Behavior, strength, and deformations observed in these two monotonic tests are discussed.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Analysis of large-scale tests for AP-600 passive containment cooling system

All next-generation light water reactors utilize passive systems to remove heat via natural circulation and are significantly different from past and current nuclear plant designs. One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 h without action by the reactor operator. During a design-basis accident (DBA), i.e., either a loss-of-coolant or a main-steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annular space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-1D code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single-phase flow, transport equations for the k − two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-1D results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized.     

Experimental interpretation and code validation based on the PHEBUS-FP programme: Lessons learnt from the analysis of the containment scenario of FPT1 and FPT2 tests

The phenomenological scenario of a severe accident is extremely complex. Its simulation requires specific models of phenomena of different nature (i.e., materials behavior, thermal–hydraulics, aerosols, etc.) and their adequate coupling in safety computer codes. Therefore, an exhaustive and extensive validation against representative databases is mandatory. PHEBUS-FP project is, beyond any doubt, one of the more valuable data sources for this purpose.The main lessons learnt from the containment simulations of FPT1 and FPT2 tests were summarized. Several safety computer codes: CONTAIN 2.0, MELCOR 1.8.5 and ASTEC 1.1 have been used. This diversity has allowed a “user-independent” cross comparison of codes and data. Overall, codes estimates have reproduced properly experimental trends and no variable has shown major discrepancies. By means of parametric studies, it has been demonstrated that the minor discrepancies found did not come from the hypotheses and approximations adopted. In addition, the analyses of codes results have assisted in the interpretation of experiments by showing potential experimental uncertainties (i.e., steam injection in FPT1) or even by crediting data from a specific measurement technique over other data sources (i.e., samplings over γ-spec in FPT1).     

Design and full scale test of the fuel handling system

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.     

Research status and needs for shear tests on large-scale reinforced concrete containment elements

Reinforced concrete containments at nuclear power plants are designed to resist forces caused by internal pressure, gravity, and severe earthquakes. The size, shape, and possible stress states in containments produce unique problems for design and construction. A lack of experimental data on the capacity of reinforced concrete to transfer shear stresses while subjected to biaxial tension has led to cumbersome if not impractical design criteria. Research programs recently conducted at the Construction Technology Laboratories and at Cornell University indicate that design criteria for tangential, peripheral, and radial shear are conservative.This paper discusses results from recent research and presents tentative changes for shear design provisions of the current United States code for containment structures. Areas where information is still lacking to fully verify new design provisions are discussed. Needs for further experimental research on large-scale specimens to develop economical, practical, and reliable design criteria for resisting shear forces in containment are identified.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

Fracture mechanics in creep situation interpretation of results on a test specimen

Fracture mechanics in creep situation is a difficult challenge for the 1990s. In France, CEA Saclay has conducted experimental tests on compact tension (CT) specimens at 650°C in order to investigate crack initiation under creep situations. The constitutive material is the 316SPH austenitic stainless steel used for most LMFR structures.Numerical simulations using SYSTUS code and simplified method analysis were performed on one of the tests (CT specimen at 650°C under constant load) to compare some parameters (notch opening, initiation time) with experimental values. The material constitutive law was represented by the complete elasto-viscoplastic CHABOCHE model for computation. Owing to geometrical characteristics such as thickness, the situation of the CT specimen was likely to be intermediate between plane stress and plane strain assumptions. From C^* parameter, incubation time obtained using the R5 rule was conservative in comparison with the test result.The continuum damage model developed at Ecole des Mines de Paris has also been used to assess creep damage in the notch tip area. The crack initiation time has been deduced from critical damage at characteristic distance (X_c = 0.05 mm). Considering critical damage specifically, for a CT specimen (D_c = 0.05), initiation time obtained was higher than the test result.The results of this study will contribute to the development of a methodology for nocivity analysis of cracks in creep situation.     

Hydraulic characterization of the full scale divertor cassette prototype

In the frame of the activities related to ITER divertor R&D, ENEA C.R. Brasimone was in charge by Fusion For Energy (F4E) to perform the assembly, the hydraulic tests and the theoretical simulation of the hydraulic behavior of the full scale divertor cassette prototype. The objective of these activities was aimed at the investigation of the thermal-hydraulic behavior of the full-scale divertor cassette both under steady state condition and during draining and drying operational transient. In particular, the steady state tests were focused on finally check whether the hydraulic design of the divertor components is able to ensure a uniform and proper cooling for the plasma facing components, with an acceptable pressure drop; whilst the transient ones were aimed at defining proper procedures for draining and drying the divertor cassette as well as for refilling it with water.This paper presents the results of the steady state and transient hydraulic experimental test campaigns performed at ENEA C.R. Brasimone as well as the relevant numerical analysis performed at the Department of Nuclear Engineering of the University of Palermo adopting the RELAP5 Mod3.3 thermal-hydraulic system code.     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.     

UPTF experiment: a synopsis of full scale test results

A series of 14 tests has been run at UPTF – a 1:1 scale test facility – to investigate the thermohydraulic phenomena in a PWR primary system during blowdown, refill and reflood phases. A synopsis of the most significant test results is given to characterize the controlling phenomena in a full scale primary system under LOCA conditions.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

Full scale dynamic tests of Atucha II NPP

This paper summarizes the main results of a series of dynamic tests of the reactor building of Atucha II NPP performed to determine the dynamic properties of its massive structure deeply embedded in quaternary soil deposits. Tests were performed under two different types of loading conditions: steady state harmonic loads imposed by mechanical exciters and impulsive loads induced by dropping a weight on the ground surface in the vicinity. Natural frequencies and mode shapes were identified and the associated modal damping ratios were experimentally determined. Numerical analyses of the reactor building-foundation system by two different F.E. models were performed. One of them, based on an axisymmetric representation of the soil-structure system, was used to simulate the steady state vibration tests and to calculate the dynamic stiffness of the foundation slab and soil layers for comparison with those experimentally obtained. The other, a 3-D F.E. model of the superstructure, was used to assess the natural frequencies and mode shapes obtained from the tests, representing dynamic stiffness of the foundation with stiffness coefficients derived both from the tests and from the axisymmetric F.E. model. Good agreement of the natural frequencies given by two types of tests were generally found, with the largest difference between them in the fundamental frequency of the building. Estimates of modal damping derived from the tests showed significant differences depending on the technique used to calculate them. For the fundamental mode, damping was found to be 23–42%, gradually decreasing with frequency to 2–4% for 10 Hz.     

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).¹ The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.     

Analysis of molten pool physico-chemical interactions and interpretation of the Phebus FP tests observations

The new 2D model for non-equilibrium U–Zr–O melt oxidation is developed on the base of the previously developed 1D model and extended to consideration of the general case of simultaneous UO₂ dissolution and melt oxidation accompanied with growth of the peripheral ceramic layer (crust) and bulk ceramic precipitates. The model is validated against isothermal crucible tests where precipitation of ceramic phase during molten Zr interactions with ceramic crucible walls along with oxide growth on non-oxidised surface of the melt were observed. In order to analyse melt oxidation under transient temperature conditions of the Phebus FP tests, the new physico-chemical model was tightly coupled with the heat exchange model and applied to interpretation of the post-test microstructure observations of the molten pool in the FP bundles.     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

Design and fabrication of a full elastic sub-micron-Newton scale thrust measurement system for plasma micro thrusters

In this work,a force measurement system is proposed to measure the thrust of plasma micro-thruster with thrust magnitude ranging from sub-micro-Newtons to hundreds micro-Newtons.The thrust measurement system uses an elastic torsional pendulum structure with a capacitance sensor to measure the displacement,which can reflect the position change caused by the applied force perpendicular to the pendulum axis.In the open-loop mode,the steady-state thrust or the impulse of the plasma micro-thruster can be obtained from the swing of the pendulum,and in the closed-loop mode the steady-state thrust can be obtained from the feedback force that keeps the pendulum at a specific position.The thrust respond of the system was calibrated using an electrostatic weak force generation device.Experimental results show that the system can measure a thrust range from 0 to 200 μN in both open-loop mode and closed-loop mode with a thrust resolution of 0.1 μN,and the system can response to a pulse bit at the magnitude of 0.1 μN s generated by a micro cathode arc thruster.The background noise of the closed-loop mode is lower than that of the open-loop mode,both less than 0.1 μN//√Hz in the range of 10 mHz to 5 Hz.     

Analytic study of the steel liner near the equipment hatch in a 1:4 scale containment model

A containment scale-model test, performed at Sandia National Laboratories, was loaded by overpressurization and the first leak was concluded to be caused by tears in the steel liner found near the equipment hatch. These tears were located in the vicinity of the vertical fold in between the general curved part and the embossment (vertical bend line). A 3D finite element analysis of the region near the equipment hatch, shows that high localized strains will develop in the vicinity of the bend line. It is shown that the liner separates from the concrete wall near the bend line when the containment expands. The tensioned liner will be in contact with the surface of the concrete wall in general, but near the vertical bend line the liner tends to be straightened out. This flexural behaviour cause high strains in the weld located in the bend line. The actual peak strain level is depending on the detailed geometry in the bend line and the failure strain level of a welded biaxial stressed zone is difficult to define. However, the analysis presented in this paper shows that the flexural behaviour in the bend line most likely contributed to the liner tears found in the scale-model test. A general conclusion from the study presented in this paper is that, the non-linear plastic behaviour of the liner is very sensitive to the detailed design and the interaction between the liner and the concrete.