Weiju Ren

A Review Paper on Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications

To understand the response of Alloy 617 to long-time exposure conditions and to determine the supplementary data needs for structural components in Gen IV nuclear reactors, literature of aging and aging effects in the alloy was reviewed. Most of the reviewed data were produced in connection with the international research effort supporting high temperature gas-cooled reactor projects in the 1970s and 1980s. Topics considered included microstructural changes, hardness, tensile properties, toughness, creep-rupture, fatigue, and crack growth. It became clear that, for the long-time very high-temperature conditions of the Gen IV reactors, a significant effort would be needed to fully understand and characterize property changes. Several topics for further research were recommended.

A Review on Current Status of Alloys 617 and 230 for Gen IV Nuclear Reactor Internals and Heat Exchangers

Alloys 617 and 230 are currently identified as two leading candidate metallic materials in the down selection for applications at temperatures above 760 C in the Gen IV Nuclear Reactor Systems. Qualifying the materials requires significant information related to Codification, mechanical behavior modeling, metallurgical stability, environmental resistance, and many other aspects. In the present paper, material requirements for the Gen IV Nuclear Reactor Systems are discussed; certain available information regarding the two alloys under consideration for the intended applications are reviewed and analyzed. Suggestions are presented for further R&D activities for the materials selection.

Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700°C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.Copyright

A Review of Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications

The literature was reviewed of aging and aging effects in Alloy 617 to determine the supplementary data needed to understand the response of the alloy to long-time exposure conditions being considered for structural components in Gen IV nuclear reactors. Most of the data were produced in connection with the international research effort supporting High Temperature Gas-Cooled Reactor (HTGR) projects in the 1970s and 1980s. Topics considered included microstructural changes, hardness, tensile properties, toughness, creep-rupture, fatigue, and crack growth. It became clear that, for the long-time, very high temperature conditions of the Gen IV reactors, a significant effort would be needed to fully understand and characterize property changes. Several topics for further research were recommended.Copyright

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Since 2002, the Department of Energys (DOEs) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOEs Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOEs structural materials research activities being conducted to support VHTR development. By far, the largest portion of materials R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

A Review of Alloy 800H for Applications in the Gen IV Nuclear Energy Systems

Alloy 800H is currently under consideration for applications in the Next Generation Nuclear Plant at operational temperatures above 750°C. To provide supporting information in this paper at the attempt to facilitate the consideration, service requirements of the nuclear system for structural materials is first described; and then an extensive review of Alloy 800H is given on its codification with respect to development and research history, mechanical behavior and design allowables, metallurgical aging resistance, environmental effect considerations, data requirements and availability, weldments, as well as many other aspects relevant to the intended nuclear application; an finally further research and development activities to support the materials qualification are suggested.Copyright

Interoperability of Materials Database Systems in Support of Nuclear Energy Development and Potential Applications for Fuel Cell Material Selection

Materials database interoperability has been of great interest in recent years for information exchange in support of research and development (R&D). In response to data and knowledge sharing needs of the GenIV International Forum (GIF) for global collaboration in nuclear energy R&D, the European Commission JRC Institute for Energy and Transport (JRC-IET) and the Oak Ridge National Laboratory (ORNL) have established a materials database interoperability project that develops techniques for automated materials data exchange between systems hosted at the two institutes MatDB Online at JRC IET and the Gen IV Materials Handbook at ORNL, respectively. The work to enable automated exchange of data between the two systems leverages the XML data import and export functionalities of both systems in combination with recently developed standards for engineering materials data. The preliminary results of data communication between the two systems have demonstrated the feasibility and efficiency of materials database interoperability, which constructs an interoperation framework that can be seamlessly integrated into the high-throughput First Principles material databases and thus advance the discovery of novel materials in fuel cell applications.

Initial Development in Joining of ODS Alloys Using Friction Stir Welding

Solid-state welding of oxide-dispersion-strengthened (ODS) alloy MA956 sheets using friction stir welding (FSW) was investigated. Butt weld was successfully produced. The weld and base metals were characterized using optical microscopy, scanning electronic microscopy, transmission electronic microscopy, and energy dispersion x-ray spectrum. Microhardness mapping was also conducted over the weld region. Analyses indicate that the distribution of the strengthening oxides was preserved in the weld. Decrease in microhardness of the weld was observed but was insignificant. The preliminary results seem to confirm the envisioned feasibility of FSW application to ODS alloy joining. For application to Gen IV nuclear reactor heat exchanger, further investigation is suggested.

A Brief Review of Models Representing Creep of Alloy 617

Alloy 617 is being considered for the construction of components to operate in the Next Generation Nuclear Plant (NGNP). Service temperatures will range from 650 to 1000°C. To meet the needs of the conceptual designers of this plant, a materials handbook is being developed that will provide information on alloy 617, as well as other materials of interest. The database for alloy 617 to be incorporated into the handbook was produced in the 1970s and 1980s, while creep and damage models were developed from the database for use in the design of high-temperature gas-cooled reactors. In the work reported here, the US database and creep models are briefly reviewed. The work reported represents progress toward a useful model of the behavior of this material in the temperature range of 650 to 1000°C.Copyright

Gen IV Materials Handbook Functionalities and Operation

This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.