High-temperature Reactor Materials Research Articles

Development of a structural integrity evaluation program for elevated temperature service according to ASME code

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

Microstructure Evaluation of Ceramic Materials Used in the Generation IV Reactors

Ceramics are candidate materials for high temperature reactors as perspective materials for fuel claddings, isolation, holders, exchangers and other heat-stressed components. In this context, particularly Al2O3 which can resist high temperatures is often mentioned as cladding material. Ceramics can also be used for high-temperature systems, mainly for components that are not directly exposed to neutron flux in the reactor core. Under this study, high alumina products - mullite Lunit 73, alumina products Luxal 203 and AG 202 were exposed to the high-temperature air atmosphere and helium atmosphere up to 900°C. Exposed specimens were mechanically polished to study the microstructural changes after exposition. Microstructural evaluation was performed by Light Optical and Scanning Electron Microscopy. Luxal 203 and AG 202 showed no changes in the microstructure, whereas in Lunit 73 new phases were observed.

MatDB Online analysis of HTR materials data from European R&D

Recent high temperature reactor (HTR) concepts for nuclear co-generation foresee hot gas temperatures up to 850°C. While suitable materials for the high temperature exposed components well-known from conventional technologies are taken into consideration, extended material qualification tests must still be performed for reactor design and licensing to guarantee safe and economic operation. For this purpose data management tools are required. For administration of materials data coming from European R&D projects the Joint Research Centre – Institute for Energy (JRC-IE) offers the web-enabled MAterials DataBase (MatDB Online) free of charge to the European research community. It represents a comprehensive tool for storing, retrieving and evaluating materials test data together with related documentation in a linked document management database (DoMa). MatDB Online hosts materials data from the former German high temperature reactor programme and was, respectively is involved in several GENeration IV (GEN IV) reactor systems related European R&D projects such as HTR-M/M1 (high temperature reactor materials) and RAPHAEL (reactor for process heat, hydrogen and electricity generation) and the cross-cutting project GETMAT (GEN IV and transmutation materials).The paper gives an overview of MatDB Online and DoMa as a central storage system for European R&D materials data and documents, describes the HTR materials and the data content within MatDB Online, and gives evaluation examples of HTR materials data. Noting the data transfer obligations within GIF (GEN IV International Forum), the paper concludes with recommendations for further materials testing.

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon–carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon–carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10 12 n/cm 2/s at temperature of 40 °C. The fluence was 2.52 × 10 16 n/cm 2. These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.

Mechanical design methods for high temperature reactor components

The work concerned with design codes for high temperature reactor (HTR) components operating at temperatures above 800°C is summarized. Using the experimental results from the German HTR materials development programmes, in particular the time dependent properties, the structural design analysis for an intermediate heat exchanger is discussed, with reference to creep, fatigue, creep buckling and creep ratchetting. The analysis provides the basis for a critical consideration of ASME Code, Case N 47, and the applicability of the code case rules for service temperatures above 800°C.

151. Development and irradiation testing of graphite and matrix materials for high temperature reactors in the Federal Republic of Germany

151. Development and irradiation testing of graphite and matrix materials for high temperature reactors in the Federal Republic of Germany

Oxidation Behavior of Nb and Nb-V Alloys

Nb and Nb-V alloys are promising as high-temperature reactor materials. For this reason their oxidation behavior was studied in the temperature range 485°-735°C. The specimens were prepared by electron-beam and arc melting. The work included kinetic studies of the oxidation as well as X-ray studies of oxide films. The oxidation reactions of Nb and Nb-V alloys were found to have a very peculiar temperature dependence. The complex oxidation behavior shown by the individual Nb and Nb-V alloys is suggested to be due to the formation of oxide layers with differing compositions in different temperature regions. The formation of a double oxide NbVO4 was identified, and the structure was examined by X-ray diffraction.