Irradiation Creep of SiC/SiC Composites. Stage 1. Development of Method and Test Facility Modernization for SiC/SiC Composites In-reactor Testing.
Tech Area / Field
8 Project completed
Senior Project Manager
Tocheny L V
FEI (IPPE), Russia, Kaluga reg., Obninsk
- Kyoto University / Institute of Advanced Energy, Japan, Kyoto
Project summaryIt is generally accepted that the development of fusion power is essential to the future advancement of human civilization, and that it is capable of exerting a relatively low impact on the environment. Although fusion power is considered to be relatively clean in comparison to fission power because there are no fission products or actinides generated, there will be a substantial amount of neutron-induced radionuclides formed in currently available structural materials.
Public acceptance and funding of fusion power can be further ensured by development of "low-activation" materials. Currently there is a strong international research focus on the removal of easily activated elements from conventional structural materials, replacing selected elements with other elements less prone to activation. These low-activation materials are not completely free of radioactivity, however, and therefore represent primarily a short-term solution only. Other less conventional materials are being sought for long-term solutions.
One of the most promising long-term solutions lies in the development of composite ceramic materials such as SiC fibers embedded in a SiC matrix, to be used both to construct the fusion reactor first wall and for construction of other important near-plasma components. The radioactivation of this composite material would be very low and would in fact be dominated by the activation of impurity elements. These impurities most likely can be controlled to produce a truly low-activation material.
The development and fabrication of suitable composite materials represents a challenge for the fusion power community, but since such a material has many potential non-nuclear applications, it is being studied in many laboratories around the world. Since such composite materials are just now being developed, there is only a very small amount of data available on their response to irradiation, a critical element for confident application of this material to fusion power.
In particular, there is only a minuscule amount of data on irradiation creep of SiC and no data on irradiation creep of SiC/SiC composite materials. Irradiation creep is one of the major processes by which the dimensional stability of structural components can be altered by radiation exposure. To date the only relevant creep data have been generated very recently on SiC fibers using charged particle irradiation.
Unfortunately, the major nations involved in SiC/SiC development and its fusion-relevant testing (USA, Japan and the European community) do not have a suitable in-reactor test facility capable of producing the required data. Under the auspices of the ITER program (International Thermonuclear Experimental Reactor) and other international collaborative fusion energy programs, however, there is a solution to this problem using research facilities and existing capabilities in Russia.
For the last 20 years the Institute of Physics and Power Engineering (IPPE) has developed and successfully used an in-pile irradiation creep facility to study the creep of stainless steels, as well as various zirconium and vanadium alloys. This facility is located in the BR-10 experimental fast reactor. The hot cell facilities of IPPE are also located close to the reactor to allow post-irradiation examination of the specimens.
It is therefore proposed that the BR-10 facility is used to perform an in-situ irradiation creep experiment on SiC/SiC composite materials and that the IPPE hot cell facilities be used to perform appropriate examinations after the irradiation is complete.
The unique and unusual nature of the SiC/SiC composite material presents a series of technological challenges for the conduct of this experiment. These challenges can be overcome by close collaboration between the IPPE staff and the various international partners. These partners are:
- Pacific Northwest National Laboratory: Ph.D. Russell H. Jones, Ph.D. Frank A. Garner.
- Oak Ridge National Laboratory: Ph.D. Arthur Rowcliffe, Ph.D. Lance Snead.
- JUPITER program participants (A large consortium of Japanese Universities performing collaborative fusion-relevant irradiations in US reactors with US researchers): Ph.D. Akira Kohyama
- National Research Institute for Metals: Ph.D. Tetsuji Noda.
- Japan Atomic Energy Research Institute: Ph.D. Akimichi Hishinuma.
- Joint Research Centre at Ispra Ph.D. P. Fenici.
Typical problems that are currently being addressed by the partners are the selection and production of the most promising candidate specimens, design and testing of the appropriate heating mode, selection of the temperature and atmosphere conditions, design and testing of suitable gripping techniques to accommodate the unusual nature of these materials, and the expansion of the current single specimen test arrangement to allow simultaneous irradiation and sequential testing of multiple specimens.
The initial experimental test matrix will also be defined jointly by the partners, but it is anticipated that the test program will be continuously modified as knowledge of the creep behavior accumulates. The effects of stress and temperature histories are particularly important. It is also anticipated that tests to failure will be made, where fracture occurs naturally as a consequence of creep, and possibly where failure is deliberately induced by strong transients in applied stress.
The Western and Japanese partners will also provide theoretical calculations to determine the displacement dose and dose rate for the SiC/SiC material. While such doses are easily calculated for conventional metal alloy components, there is, at mis time, no well-defined procedure for such complex ceramic materials.
The international partners will supply some of the test material and may manufacture the specimens and some gripping components. The partners will also participate in the on-site testing and post-irradiation examination. It is envisioned that a three year program would include approximately 6-8 months to modify and test currently existing facilities, 16 months of irradiation with continuous data acquisition, followed by 6 months or more of post-irradiation examination, analysis and joint publication of results.
The funding supplied by ISTC will be devoted primarily to support of the IPPE personnel required to run and maintain the reactor, and to conduct the tests and the post-irradiation examination. No funds will be applied to "neutron charges", as is typical of tests conducted in Western reactors.
The successful conduct of this program may have the additional benefit of attracting other creep programs funded by outside programs. This possibility will become particularly important as the test reactors in the Western countries continue to be shut down and new materials continue to be developed.
Reactor allowance including power supply, guard, overheads and experimental rig decommission, including radioactive waste utilization, are the IPPE investment into irradiation creep of SiC/SiC. The cost of this investment is estimated about 500,000 US$.
To assist in the development of future projects beyond the current scope, the program will seek to integrate IPPE staff into on going activities related to fusion materials, irradiation creep and especially ceramic composite materials. Therefore, two international trips by IPPE staff each year to visit and participate in related activities of the partners countries are proposed. Also, an international workshop on SiC/SiC and their behavior will be held at IPPE. The IPPE scientists will also be invited to visit the various participating laboratories and universities.
It is also envisioned that the international partners will visit IPPE, with some serving assignments of 2-4 weeks or longer to participate in the program. The lead US researcher, F.A.Garner, has already made four trips to IPPE in 1996-1997 to define the program.
While F.A.Garner will serve as nominal leader for the international partners, each participating group will have an equal voice in defining the direction of the program. A.Kohyama will serve as point of contact for the Japanese participants and P.Fenici for the EC
Brief Discussion of Test Technique and Program
A currently available creep facility located in the OK-70 channel of the BR-10 reactor will be used. This channel is of tubular shape and operates in a dry mode (vacuum or helium atmosphere). The channel is ~70 mm in diameter and possesses a fast neutron flux of about 6.9 ґ 1013n ґ cm-2 ґ s-1, which would produce ~1 dpa/year in metals. The corresponding dpa rate for SiC will be significantly larger but is currently not well defined.
The test temperatures will range from 600 to 800° C, with temperatures controlled by a combination of gamma heating, electrical heating and atmosphere control. The creep measurement technique will utilize the stress relaxation/stress reapplication technique previously employed at IPPE. Three flat specimens of different length will be placed in a parallel arrangement, with stress applied to only one specimen at a time. The other specimens will initially receive irradiation in absence of stress to simulate one important aspect of potential fusion service. The first specimen will be tested to natural failure or will be deliberately broken, depending on its prior behavior. The next specimen will then be loaded and tested in a similar manner, followed by a test on the third specimen, each subjected to relevant stress and temperature histories.
The majority of the equipment and software for data compilation and manipulation is already in place with only minor alterations and upgrades anticipated.
A full range of post-irradiation examination techniques will be used to study radiation-induced microstructure, fracture mode and topography, and the mechanical properties of the failed specimens. If one or two of the specimens survive without failure, ex-reactor tensile tests may also be conducted.