Taiju Shibata

Development of an Evaluation Model for the Thermal Annealing Effect on Thermal Conductivity of IG-110 Graphite for High-Temperature Gas-Cooled Reactors

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.

Interpolation and Extrapolation Method to Analyze Irradiation-Induced Dimensional Change Data of Graphite for Design of Core Components in Very High Temperature Reactor (VHTR)

Graphite materials are used as core components in the High-Temperature Gas-Cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). The authors prepared technical documents for design, material, products, in-service inspection and maintenance of the graphite components for the HTGR/VHTR, which were summarized as a draft of standard for the graphite components through discussion made in a “Special committee on research on preparation for codes for graphite components in HTGR” set up within AESJ. The draft of standard contains graphical expressions for the irradiated material properties of IG-110 graphite. It is possible to use the graphical expressions for the components design of VHTR. The graphs were obtained based on the interpolation and extrapolation of the irradiation data. The irradiation-induced dimensional change of IG-110 graphite was obtained through the interpolation and extrapolation of the irradiation data with a quadratic equation of fast neutron fluence. The irradiation data for H-451 and ATR-2E graphites were used for the evaluation of the interpolation and extrapolation of irradiation data for IG-110. It was shown in this study that the proposed interpolation and extrapolation method is reasonable for IG-110 with regard to the database available at present.

Investigation of Microstructural Change by X-ray Tomography and Anisotropic Effect on Thermal Property of Thermally Oxidized 2D-C/C Composite for Very High Temperature Reactor

Two-dimensional carbon fiber reinforced carbon composite (2D-C/C composite) is one of the candidate materials for reactor internals, e.g., control rod element, of Very High Temperature Reactor (VHTR) because of its high strength at high temperature and thermal stability. From the viewpoint of its application to the reactor internals of the VHTR, it is important to investigate the anisotropic effect on its properties for the design and safety analysis of the VHTR. Since the properties of the 2D-C/C composite are strongly dependent on its microstructure, it is necessary to observe its microstructural variations to correlate to the changes in its properties. This study has shown that X-ray tomography can be applied to observe the internal microstructural change of the thermally oxidized 2D-C/C composite. The relationship between the change in properties, including the thermal conductivity, coefficient of thermal expansion (CTE), and burn-off of the thermally oxidized 2D-C/C composite, can be expressed using the empirical exponential decay formula in both directions perpendicular and parallel to the lamina. The direction of the hexagonal graphite crystal structure from the carbon atoms and the microstructure of the 2D-C/C composite can explain not only the relationship between changes in the thermal conductivity, CTE, and burn-off but also the difference in the changes in the thermal conductivity and CTE between fiber directions.

Investigation on structural integrity of graphite component during high temperature 950 °C continuous operation of HTTR

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950 °C continuous operation, a high temperature continuous operation with reactor outlet temperature of 950 °C for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950 °C continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000 °C. In addition, the programs of surveillance test and ISI using TV camera were introduced.

Study on Fracture Behavior of 2D-C/C Composite for Application to Control Rod of Very High Temperature Reactor

For a control rod element of the Very High Temperature Reactor, a carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials for its high strength and thermal stability. In this study, in order to establish the data base of the 2D-C/C composite, the fracture data was obtained by simulating the crack expected to be generated under the VHTR condition and the oxidation effect on the fracture behavior was evaluated. Moreover, the fracture mechanism of the C/C composite was investigated through scanning electron microscope observation. This study showed that the oxidized matrix caused reduction of the fracture toughness and the reduction ratio was dependent on the density of matrix and a number cracks. With increasing the oxidation, the fracture toughness is mainly dependent on the fiber characteristics. Furthermore, the crack grows along the boundary between fiber bundles without breaking the fiber. The cracks which were initiated at the interface between the matrix and the fiber were gathered into the voids in the boundary between fiber bundles, and, then, the cracks grew up in the matrix.

Research and developments on application of carbon-carbon composite to HTGR/VHTR in Japan

High Temperature Gas-cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR) are attractive nuclear reactors to obtain high temperature helium gas at the reactor outlet. To enhance the thermal efficiency, the in-core internals of HTGR/VHTR, especially control rods, are subjected to the severe thermo-mechanical condition. The carbon-carbon (C/C) composite is one of the advanced material candidates for the control rod sheath of the advanced reactors where the excellent thermal resistance and stability are required because of the possible severe condition. The Research and development on the C/C composite application to HTGR have been carried out since 1990s. JAEA and Toyo Tanso have carried out the R & D on C/C composite to be used for control rod. Application of C/C composite is recently focused as one of the important subjects to develop VHTR in the international R & D activities. Scheme of the development in the JAEA/Toyo collaboration is outlined as follow: After the feasibility of C/C composite rod was demonstrated by a conceptual design, the procedure is progressing as follows; (1) Database establishment, (2) Design and manufacturing of components, and (3) Demonstration test by High Temperature engineering Test Reactor.

Correlation of Microstructure and Compressive Strength of C/C Composite Using X-ray Tomography

For the control rod element of a Very High Temperature Reactor, carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials for its high strength and thermal stability. In this study, in order to correlate the microstructure of the C/C composite to its compressive strength, the X-ray tomography was applied to visualize the internal microstructure of the C/C composite. The relationship between change in the compressive strength and that in the microstructure was also investigated. This study showed that the pore distribution in the C/C composite could be confirmed visually and the volume and shape of the pores could be evaluated by the X-ray tomography in three-dimension. Moreover, since the matrix was gradually lost and transverse cracks became large with increasing the oxidation, the bonding strength between fiber bundles became weak and the compressive strength of parallel to lamina decreased.

Core Design Study of Small-Sized High Temperature Reactor for Electricity Generation

A High Temperature Gas-cooled Reactor (HTGR) has several features different from conventional light water reactors such as inherent safety characteristics, high thermal efficiency and high economy. On the other hand, one of disadvantages of the HTGR with a prismatic core is to require rather long-term and expensive refueling, resulting in relatively long maintenance period and high cost. To solve the disadvantage, the present study challenges the core design of a small-sized reactor for long refueling interval by increasing core size, fuel loading and fuel burn up compared with the High Temperature engineering Test Reactor (HTTR). The preliminary burn-up calculation suggested that approximately 6 years of long refueling interval was found to be reasonably achieved. A refueling interval longer than 6 years may be possible by decreasing further power density, subsequently larger core size with operational reactor power of 120MWt, but this idea was not taken by the requirement of the reactor that the core size shall be accommodated reasonably in the core with double size of the HTTR at maximum.Copyright

Analytical Study on Micro-Indentation Method to Integrity Evaluation for Graphite Components in HTGR

An analytical study on micro-indentation method to integrity evaluation for graphite components was carried out. The indentation method is used as simplicity test to measure mechanical properties of materials. This method is thought to be applicable to evaluate the residual stress from the relationship between indentation load and indentation depth. In this study, in order to confirm the applicability of the micro-indentation method for lifetime evaluation of the graphite component, indentation load-depth behavior under stress/strain condition was evaluated taking account of the specified minimum ultimate strength of IG-110 graphite. Moreover, analytical investigations of indentation load-depth behavior for oxidized graphite and oxidized graphite with residual strain were also carried out. As a result, it can be said that the indentation method is potentially applicable to evaluate the integrity of graphite components. (authors)