Evaluation of aseismic integrity in the HTTR core-bottom structure V. On the static and dynamic behavior of graphitic HTTR key-keyway structures

The graphite components in high temperature gas-cooled reactors are connected to each other through a key-keyway structure that has gaps between the key and the keyway to accomodate thermal expansion. Because a dynamic load concentrates on the key-keyway structure during earthquakes, it is considered to be a crucial element for assessing the integrity of the graphite components. A combination of experiments and analyses was employed to investigate the dynamic behavior of the key-keyway structure, i.e. the equivalent stiffness associated with vibrational characteristics of the graphite components and the stress distribution under dynamic loading. The experiments were performed using a graphite scale model and a dynamic photo-elastic method. The analysis was carried out using the finite element method (FEM) code Abaqus, taking account of the contact between the key and the keyway. The following conclusions were derived. (1) The equivalent stiffness of the key-keyway structure shows nonlinearity, owing to the contact deformation. (2) The equivalent stiffness evaluated by the FEM analysis, taking account of the non-inear contact deformation, is applicable for predicting the vibrational characteristics of ky-keyway structure. (3) The stress concentration under dynamic loading is lower than or nearly equal to that under static loading. The maximum stress concentration of the seismic load can be sufficiently evaluated under static loading conditions.     

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

A commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.

The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.     

Review on condensation on the containment structures

In the last two decades condensation on the containment structures in presence of noncondensables has received substantial attention by nuclear scientists and engineers. The reason is that many Generation III and III+ reactors rely on passive systems operating under natural circulation. As a consequence, a vast number of publications have been made in the open literature. This paper reviews the specific physical phenomena that are involved in condensation and discusses how they have been considered in the different available models. In addition, it explores the data that have been used for validation and provides some insights on the effective suitability for this purpose. Finally, the paper summarizes the current codes' capabilities to deal with wall condensation in the presence of noncondensables according to the most recent available validation exercises.     

An overview of reactor containment structures

The purpose of this paper is to present an overview of reactor containment structures and to summarize the present state-of-the-art of containment design. The areas covered are types of containments used for nuclear power plants in operation and under construction, and their development. Also presented are codes which currently govern the design, materials, and construction of containments, as well as some thoughts on safety and methods of analysis.     

Reliability assessment of reinforced concrete containment structures

A reliability analysis method for seismic category I structures subjected to various load combinations is developed and numerical examples are worked out under various assumptions and idealizations. The method falls generally within the so-called level III category within the framework of reliability analysis and design.     

Penetration resistance of reinforced concrete containment structures

Containment structures not only provide a leak tight barrier, but also play a role in ensuring that the structures can withstand the impact load from projectile impacts or internal plant accidents. In assessing the containment structures of nuclear power plants, predicting the characteristics of impact resistance in relation to design and safety considerations is relevant. This investigation proposes a simple but effective method of performing numerical analysis on perforation resistance of reinforced concrete containment structures. In this work, normal and oblique impacting is considered to examine the residual velocity and impact phenomena of an ogive-nose steel projectile with various impact velocities against a reinforced concrete slab. Additionally, a phase diagram is devised to describe the ballistic terminal phenomena of projectile and target. This model could assess the resistance to penetration to results in the optimum design of the containment structures in nuclear power plants.     

Some aspects of the reliability-based design of reactor containment structures

A rational procedure for the design of reactor containment structures is carried out within a probabilistic framework. Various risk concepts such as the return period, non-encounter probability and the reliability function are discussed. Internal load conditions caused by system failure such as LOCA pressure loads, and external load conditions caused, for instance, by impact due to aircraft crashes, external pressure waves and natural hazards such as earthquakes and severe storms, are described by extreme value distributions of the largest values of the Fisher-Tippett types. Statistical and physical arguments are given to support their application. The occurrence of these rare events with respect to time is modeled by a Poisson process. The ultimate strength of a PWR containment structure for the steel (liner) shell is also modeled by an extreme value distribution (of the smallest values). As a good approximation the load action of the shell structure is determined by linear elastic analysis. The failure criterion considered here is that of reaching the ultimate tensile strength at one point of the structure. A numerical example of the reliability analysis of a steel shell structure under internal overpressure is carried out.     

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.     

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.     

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.     

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.     

Theoretical study of aircraft impact on reactor containment structures

In containment design there is a requirement to protect the reactor system from the effects of external hazards and hence it is necessary to provide suitable wall thicknesses. Experimental work undertaken by the UKAEA is being carried out as a general study and this paper describes some theoretical studies for the particular case of an aircraft impact. The theoretical study utilizes a finite difference dynamic code based upon dynamic relaxation initially developed for static problems The code models concrete, reinforcement and prestressing throughout the short term non-linear range. Concrete is assumed to have a limited tensile stress capacity, coupled with a shear carrying capacity which is dependent upon the aggregate and crack size. In addition a yield condition can be specified to allow for triaxial stress states both initial and subsequent to failure. The paper briefly describes the theory and makes comparisons for different concrete thicknesses.     

900 MW PWR containment mechanical behavior characteristics during containment test

Containment mechanical behavior characteristics are one of the important checking items during 900 MW PWR containment test. This article describes these characteristics (including strain, displacement and temperature influence) systematically. Especially some intrinsic characteristics are discussed and some new viewpoints are provided. We believe that it has some reference value to other containment test.     

Reactor technology development under the HTTR project

JAERI is developing HTR technology, hydrogen production technology, and system integration technology under the HTTR Project. The HTTR is the Japanese first HTR with a 30-MW thermal power. The first criticality of the HTTR was achieved in 1998, and the full-power operation at an outlet coolant temperature of 850°C was attained in 2001. The outlet coolant temperature was reached to 950°C in 2004. A seven-year program on the gas turbine HTR was launched in 2001. The program consists of the design of a GTHTR300 plant and R&D on a closed-cycle helium gas turbine system for the GTHTR300. It is designed to have a 600-MW thermal power at an outlet coolant temperature of 850°C and a 275-MW electric power. The objectives of the program are to establish a feasible plant design and to demonstrate key technologies for the helium gas turbine. The GTHTR300 design will demonstrate competitive economy and high degree of safety. It will also provide technology basis of VHTRs for power generation, hydrogen production, and cogeneration.     

Reliability-based assessment of integrity of bonded prestressed concrete containment structures

The main function of a nuclear containment structure is to prevent the leakage of radioactive materials from the reactor in the event of a serious failure in the process system. To maintain a high level of leak integrity, prestressed concrete is widely utilized in containment construction. In bonded prestressing systems, excessive prestressing losses caused by unexpected material deformations and degradation of tendons could result in the loss of leak integrity under an accident. To safeguard against this, the Canadian Standard, CSA N287.7 (1995), recommends periodic inspection and evaluation of prestressing systems of CANDU containments. As bonded tendons are not amenable to direct inspection, the evaluation is based on the testing of a set of beams with features identical to the containment. The paper presents a quantitative reliability-based approach to evaluate the containment integrity in terms of the condition of bonded prestressing systems. The proposed approach utilizes the results of lift-off, destructive, and flexural tests to update the probability distribution of prestressing force, and to revise the calculated reliability against through-wall cracking of containment elements. An acceptable criterion for the results of beam tests is established on the basis of maintaining adequate reliability throughout the service life of the containment.     

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.     

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.     

Simplified method for evaluating the seismic buckling capacity of unstiffened steel containment structures

A simplified method is presented for evaluating the seismic buckling capacity of unstiffened, free-standing steel containment structures. The method is consistent with current US Nuclear Regulatory Commission seismic design standards and with containment buckling interaction equations given in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code which includes the influence of geometrical imperfections of the shell on buckling. Stresses to be considered in the interaction equations are determined from beam theory using standard response spectrum analysis. An empirical correction factor is developed to account for hoop stresses that are not explicitly represented in the beam theory. As the results of these analyses are very sensitive to the damping that is assumed, the extensive three-dimensional finite element analyses that were performed to develop the hoop stress reduction factor were also used to study the sensitivity of containment buckling to the assumed damping. Experiments on model containment structures were then performed to further investigate the damping properties exhibited by these structures. The study in concluded by showing that the simplified method reasonably predicts seismic buckling capacities when compared with independently determined predictions from detailed finite element analyses.