Data on Steel Irradiated in Fast Reactor
Application of BN-350 Data on Ferritic-Martensitic Steel to International Efforts on Advanced Reactor Concepts (Fusion, Accelerator-Driven, Generation-4)
Tech Area / Field
- FIR-MAT/Materials and Materials Conversion/Fission Reactors
- FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors
3 Approved without Funding
National Nuclear Center of the Republic of Kazakstan / Institute of Nuclear Physics, Kazakstan, Almaty
- MAEC-Kazatomprom, Kazakstan, Aktau
- University of Tennessee, USA, TN, Knoxville\nLos Alamos National Laboratory / Advanced Fuel Cycle Initiative, USA, NM, Los-Alamos\nPacific Northwest National Laboratory, USA, WA, Richland\nUniversity of Wisconsin-Madison / College of Engineering, USA, WI, Madison\nSCK-CEN, Belgium, Mol
Project summaryAt present in many industrial countries contribution of nuclear power to economy is quite sizable. However, its development is artificially hampered, generally, as a result of negative public opinion vested after heavy accidents in Chernobyl and Tree Mail Island. Nevertheless, the permanently increasing demands of new power capacities have stimulated adoption of the forth-generation reactor development & construction program in USA and Russia (INPRO). With public opinion taken into consideration, in these and similar programs special attention is paid to safety items reasonably combined with conserved price competitiveness of the power produced in nuclear power plants. Prospecting trends - hybrid reactors and thermonuclear facilities – are, still, under development, and their successful realization depends, in particular, on results of systematic studies of the radiation effects responsible for material serviceability (embrittlement, swelling, creep, etc.).
Up to date, the studies carried out with austenitic and ferritic stainless steels have shown that the ferritic-martensitic grade steels, compared to austenitic ones, are more stable to radiation swelling and creep at the fluence ranging up to 50 to 80 dpa. Unfortunately, loss of interest to nuclear power in the nineties made it impossible to clear up to full extent potential capabilities of ferritic-martensitic steels as constructional material for prospecting new-generation reactors.
Today a problem in choice of radiation-resistant reactor materials consists, basically, in insufficient scope of experimental data characterizing behavior of stainless highly doped steels (including ferritic-martensitic ones) at the temperature/strength/radiation fields, in particular, for various neutron spectra and parameters of the collected gas admixtures typical for fast reactors and the proposed prospecting nuclear power facilities.
In Kazakhstan, after long-term operation of the pioneer industrial fast reactor BN-350, considerable amount of spent fuel assemblies (FA) as well as units and parts made of ferritic-martensitic steels were accumulated. Comprehensive investigation of these materials, irradiated up to high damaging doses (50 to 100 dpa) and documented thoroughly, will be useful in view of solving a number of radiation material science problems, such as elimination of scientific faction related to an extent of radiation swelling in ferrite steels.
This project is aimed at execution of comprehensive material science studies and characterization of ferritic-martensitic steel 12Х13М2БФР (EP-450) irradiated in the BN-350 reactor as the FA structural material within the temperature range 280 to 390oC up to the damaging doses 80 to 90 dpa.
In the BN-350 reactor laboratory (Aktau) plates will be cut out of the spent hexagonal FA shroud s corresponding to various distances from the reactor core center. Then the plates will be transported to INP (the Alatau settlement), where they will be subject to careful examination in hot cells, profilometry and splicing into specimen for subsequent complex material science studies. In course of specimen preparation, techniques for optical metallography, scanning/transmission electron microscopy, precision mechanical trials, microhardness measurement and local microelement analysis will be applied. Also a new technique for mechanical trials - “Shear-Punch” – and, in view of making experiments on structure thermal evolution – differential scanning calorimeter – will be used.
Two Kazakhstani institutions are the project participants: the Institute of Nuclear Physics (INP) and the Mangyshlak Energy Complex (MAEC), earlier executed successfully the ISTC projects K-172 and K-437. Experienced specialists, earlier involved in radiation damage studies and creation of novel constructional materials for the transport and defense-purpose industrial nuclear power plant reactor cores, will be attracted to project realization. . Participation of these people will provide high scientific level of studies to be fulfilled.
The PNNL, Argonne, Oak Ridge and los Alamos National Laboratories (USA) are interested in the project outcomes. Collaborators from mentioned above institutions are assumed to take part in discussion of programs, work plans and project outputs.
The project is adequate to ISTC tasks and goals: it makes it possible to re-orient weapons scientists and engineers to peaceful activities; it promotes integration of Kazakhstani scientists to international scientific community, supports fundamental and applied studies in the area of nuclear safety and radiation material science. The expected results are as follows:
- Data on the structure parameters, values of the irradiated ferritic-martensitic steel mechanical characteristics versus the irradiation temperature, the fluence and the neutron flux will be obtained.
- Defect structure evolution and peculiarities in deformation behavior of irradiated ferritic-martensitic steel under uniaxial tension and Shear – Punch mechanical testing will be studied thoroughly.
- The database created as a result of these studies will be used, in particular, in course of designing the forth-generation reactors. Besides, these data are assumed to be used when developing recommendations on prolonging the light-water reactor lifetimes, as low entrance temperature of sodium (280oC) in the BN-350 reactor as well as levels of neutron fluxes and integral doses correspond to the values typical for light-water reactors.
- The project outcomes will be of high fundamental significance, because qualitatively new data on kinetics of defect structure formation in this grade of materials, diffusion processes, pore/phase formation in solid solution will be obtained.
It should be marked that the program is assumed to be realized in close matching to similar programs executed in USA, Russia and Ukraine rather than in isolation. One of main tasks of the program integrating a number of projects executed at the same time is to integrated the INP personnel, as well as Kazakhstan as a whole, along with the resources used, to international scientific community and to use fully their potential as soon as possible.
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.