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Fast Reactor Materials Properties


Application of Structural Materials Data from the BN-350 Fast Reactor to Life Extension of Light Water Reactors

Tech Area / Field

  • FIR-MAT/Materials and Materials Conversion/Fission Reactors
  • FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors

3 Approved without Funding

Registration date

Leading Institute
Nuclear Technology Safety Center, Kazakstan, Almaty

Supporting institutes

  • MAEC-Kazatomprom, Kazakstan, Aktau\nNational Nuclear Center of the Republic of Kazakstan / Institute of Nuclear Physics, Kazakstan, Almaty


  • EDF - Electricitè de France, France, Clamart\nFramatome ANP GmbH, Germany, Erlangen

Project summary

It is well known, neutron damage to the permanent internal structures of light-water (LWR) reactors is one of the limiting factors in the usable life of these reactors. Neutron damage causes embrittlement and may lead to swelling of the various grades of stainless steel from which components like the core barrel are manufactured. Although neutron damage to LWR components can be simulated by various ion bombardment techniques, there is always some modeling extrapolation necessary, so that there is no substitute for direct measurement on irradiated specimens. Unfortunately, there are very few reactor facilities world wide for specimen irradiation and, when available, such irradiations take about the same time to achieve end-of-life neutron doses as in actual LWRs. Similarly, the alternative method of obtaining samples from the permanent internal structures of decommissioned plants is extremely difficult and expensive. In consequence, the case for life extension of pressurized water reactors (PWRs) and the Russian equivalents (VVERs) is presently made on the basis of extrapolating trends observed in lower dose materials.

In the framework of ISTC K437 Project it was suggested to investigate the materials irradiated in BN-350 reactor keeping in mind the low inlet temperature of BN-350 (280 C) which means that the irradiation conditions in-reactor (dose rate, dose and temperature) bound the values occurring in LWRs and to use these data for predicting behavior of LWR internals after its operation within long period of time.

ISTC Project K-437 commenced in June 1, 2001 and completed in September 1, 2005. When implementing this project the main attention was paid to investigations of the changes in structure and mechanical properties of the steels used as the ducts of BN-350 reactor Fuel Assemblies (FA) depending on neutron flux / fluence and irradiation temperature. Project objective was an obtaining data on BN-350 reactor materials structure and properties changes for extension of the lifetime of Light Water reactors of commercial nuclear power plants.

Three Kazakhstani institutions were participated in the Project: Nuclear Technology Safety Center (NTSC), MAEC Kazatomprom, Institute of Nuclear Physics of National Nuclear Center (INP NNC RK).

Scope of works done under K437 Project included implementation of neutron-thermal calculations for FA irradiated in BN-350 reactor, selection of the objects for investigation, cutting off samples for investigation, transportation of samples from Aktau to Almaty, investigation of the irradiated samples ductility / strength changes, investigation of the irradiated materials swelling, determination of gaseous impurities content in the irradiated materials.

Samples of 08Cr16Ni11Mo3 (AISI 316 analogue) and 12Cr18Ni10Ti (AISI 321 analogue) austenitic stainless steels cut off from the different places of chosen fuel assembly ducts irradiated in BN-350 reactor have been investigated by means of optical metallography, TEM and SEM, X-ray analysis, mechanical tensile tests, microhardness measurements, density measurements, thermodesorption spectroscopy.

The most interesting results obtained in the course of K-437 ISTC Project implementation, which relate to the life extension of light water reactors are as following:

  • The voids were found in both 08Cr16Ni11Mo3 and 12Cr18Ni10Ti steels irradiated at rather low temperature 280 C up to only 1.3 dpa damage dose at dose rate 3.9×10-9 dpa/s and 0.65 dpa at 1.2×10-9 dpa/s respectively. Previously, there was a common opinion that 305 C was a threshold temperature for pore formation in austenitic stainless steels and these steels swelled only at higher temperatures. The data obtained show that pores could be found in the internals of light water reactors operated at similar irradiation conditions.
  • It was found that voids with the diameter 10-15nm and density 1÷2×1021void/m3 are nucleated in 08Cr16Ni11Mo3 austenitic stainless steel irradiated at T=302÷311 C up to damage doses 7÷13 dpa with dose rates 2.2÷3.3×10-8dpa/s, while voids were not found in the austenitic steel samples irradiated within the same temperature range up to the same damage dose but with dose rates 19÷26×10-8dpa/s. 12Cr18Ni10Ti austenitic stainless steel demonstrates similar behavior.
  • Simple model has been developed to explain the dose rate dependence of swelling in austenitic stainless steels.
  • There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7-13 dpa at 302÷311 C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. This effect was not found for 12Cr18Ni10Ti steel samples. But for both steels, there was found an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the voids formation in the steels at low dose rates. This is an important issue for light water reactors internals because, possibly, there was no previously an expectation to find a significant lost of ductility in light water reactors internal components irradiated at relatively low temperatures up to low doses and at low dose rates.
  • There was found unexpected degradation of mechanical properties of 08Cr16Ni11Mo3 steel irradiated up to damage dose 11 dpa only at dose rate 2×10-7 dpa/s and irradiation temperature 346 C. Presumably, steel properties degradation was caused by a creep effect under mechanism of grain boundary sliding.

This proposal suggests to extend K437 Project within 48 months with additional funding to perform additional investigations of austenitic stainless steels irradiated in BN-350 at low dose rates. During this proposed project extension an intention will be focused on extraction and, then, investigation of BN-350 internal component samples made of Cr18Ni9 (AISI 304 analogue) and Cr18Ni9Ti (AISI 321 analogue) austenitic steels irradiated during more than 25 years up to ~1÷10dpa damage dose within temperature range 280÷360 C. Additionally there will be carried out the investigations of fuel assembly ducts made of 08Cr16Ni11Mo3 austenitic steel which demonstrated unexpected degradation properties after being irradiated in BN-350 reactor followed by some wet storage in reactor pool, in order to clarify if such a behavior is an issue for internals of LWRs. The data obtained in the course of this material research could be used for the validation of numerical simulations of irradiation damage in LWR internal components performed in the framework of the European Integrated Project PERFECT. Under the support of European collaborators the results of the extended project could be connected to sub project II of PERFECT "RPV & Internals: Physics Modeling" WPII-6 "Internals: Long term evolution of irradiation induced damage" as well as to the work package INTERN-2, a virtual reactor for internals, "IASCC module" of PERFECT. There is an understanding that PERFECT is already running and will be finished in approximately two years from now by the time when BN-350 internals were scheduled to be extracted under the proposal. Therefore, more likely, there would not be a direct input of data from K437 Project into PERFECT at the development stage, but model calculations from PERFECT could be verified using data obtained later from the suggested ISTC K437 Project extension proposal.

Besides that, these data could be directly applied to the assessment of long time operational safety of BN-600 Russian reactor. One of the EU projects, which provides assistance to Russia through TACIS program relates to an improvement of BN-600 safety, data gained in the course of investigations suggested could be considered as supplemental ones for forecasting long time operational reliability of BN-600 internal components.

The proposed scope of works will naturally pide into several distinct tasks. In the first task irradiation conditions will be calculated for candidate internal components to establish the axial distribution of irradiation temperature, and the axial distribution of dpa/sec and fluence. MAEC Kazatomprom and NTSC will perform this task. To extract the pieces of BN-350 internal components a Design Documentation Package including Safety Evaluation Review will be developed by designers of MAEC and NTSC in the framework of second task. It is planned to extract the hexagonal elements of BN-350 core support structure (neutron shield) made of austenitic Cr18Ni9 (AISI 304 analogue) and Cr18Ni9Ti (AISI 321 analogue) steels, about three meters length and diameter ~100mm. The third task will involve getting approval of Design Documentation Package from the Kazakhstan regulatory authorities. After getting approval of Design Documentation Package by all involved parties a fabrication package for the equipment needed to extract BN-350 internals will be prepared in the framework of fourth task. The fifth task includes procurement and fabrication of necessary equipment, its installation and checkout. The sixth task will involve extraction of chosen pieces of BN-350 internal components. Implementation of this task will require an operation of rotation plugs which are currently sealed by IAEA. To start operation of BN-350 rotation plugs it will be necessary to obtain permission from IAEA to remove the seals. These negotiations will be carried out in the framework of sixth task as well. The seventh task will involve machining the pieces of internal components and cutting out specimens from them at appropriate locations in the hot cells of MAEC Kazatomprom. The eighths task will involve the packaging and transportation of specimens between Aktau and the INP NNC facilities in Alatau, Almaty; this task will be shared between MAEC Kazatomprom and INP NNC to ensure specimen activities (sizes) are compatible with the INP NNC facilities. The ninths task will be devoted to investigation of the mechanical property changes as a result of reactor irradiation. MAEC Kazatomprom will perform «preliminary» mechanical tensile tests, while INP NNC will perform «precision» mechanical tensile. Besides, optical metallography and electron microscopy will be applied. Additionally, phase-structure investigations by means of X-ray analysis will be performed. The tenth task will involve determination of swelling in the samples and will be performed in MAEC Kazatomprom and INP NNC laboratories. This task includes sample preparation for density measurements and electron microscopy, and also implementation of corresponding investigations. The eleventh task will determine the gas content of samples using thermal desorption spectroscopy; this task will be performed by NTSC specialists. In the twelfth task specialists of INP NNC, NTSC, MAEC Kazatomprom will analyze results for further development of existing model, which describes the dose rate dependence of swelling in austenitic stainless steels. In parallel to the activity described above the investigations of the ducts which demonstrated unexpected degradation of the properties will be implemented at INP NNC RK facilities under thirteenth task.

NTSC will provide both the management of the project and the necessary communications links between organizations to assure that the safety and procedural concerns of all parties were met.

ISTC Contact Expert Group on Plant Life Management (ISTC CEG PLIM) has already supervised activities and results obtained under K437 Project. This kind of mutually fruitful collaboration will continue in the future. EdF (France) and Framatome ANP GmbH (Germany) will participate in implementation of these activities as collaborators. Foreign collaborators and ISTC CEG PLIM will monitor project on a regular basis including recommendations and suggestions in the implementation of the Work Plan.


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.


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