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Fuel Assembly Materials under Dry Storage


Behavior of Nuclear Reactor Fuel Assembly Materials during Their Long-Term Dry Storage

Tech Area / Field

  • FIR-MAT/Materials and Materials Conversion/Fission Reactors

8 Project completed

Registration date

Completion date

Senior Project Manager
Valentine M

Leading Institute
National Nuclear Center of the Republic of Kazakstan / Institute of Atomic Energy (1), Kazakstan, Almaty

Supporting institutes

  • National Nuclear Center of the Republic of Kazakstan / Institute of Nuclear Physics, Kazakstan, Almaty


  • Los Alamos National Laboratory, USA, NM, Los-Alamos

Project summary

Kazakhstan has made a decision on transportation of spent fuel assemblies of the BN-350 fast reactor from Aktau-city to the former Semipalatinsk Test Site, where these are supposed to be stored for 50 years. A dry storage facility is planned for construction. At present various designs for storing fuel assemblies are under consideration. All of them provide for using sealed fuel claddings as the first barrier that preclude release of fission radioactive products into the environment. Fuel elements are emplaced in stainless steel casks that serve as the second protective barrier precluding radionuclides release. The casks with fuel assemblies are installed inside a carbon steel casing buried in the soil. Upon preliminary calculations, the temperature of fuel assembly materials in such structures may reach 400С due to decay heating. Therefore, it is important to make a long-term prediction for degradation level of the protective barrier and an estimation of the safety condition changes that may occur during the long-term storage of the spent fuel assembly materials.

The fuel assembly behavior under conditions of the long-term dry storage may be predicted using design data obtained by means of the computer codes. Their primary drawback is that data for non-irradiated and nonaged materials were used as basic design parameters. These calculations do not consider correctly a factor of continuous changing of such important properties as corrosion resistance, and mechanical properties under irradiation and dry storage conditions. During neutron irradiation of the stainless steels considerably changes its properties. Presence of the helium accumulated as a result of (n, ) reactions, may additionally effect on these changes. If we experimentally determine an activation energy of the corrosion process, we may predict the material behavior during a long-term thermal aging. However, the behavior of fuel assemblies under a long-term dry storage may be correctly predicted only if the alteration of material corrosion resistance during the storage itself will be taken into account.

The objective of this project is to predict a degradation level of the barrier materials that prevent release of fission products into the environment during 50-year dry storage of the BN-350 fuel assemblies using experimental data of the detailed investigation conducted to study regularities in the change of corrosion and mechanical properties of irradiated and non-irradiated materials during long-term thermal aging.

The project provides for performing thermalphysic calculations to determine temperature fields in the nuclear reactor fuel assemblies under conditions of dry storage, thermal aging of irradiated and non-irradiated protective barrier materials, microstructure studies of aged materials, their elemental and phase analysis; provides for measuring the density, and conducting corrosion and mechanical tests. A greater part of the work is related to a theoretical analysis of the experimental data in order to correctly predict alteration of the material properties during a long-term dry storage.

The following eight tasks are set in the project:

1. Calculation of temperature distribution in a BN-350 fuel assembly under the dry storage conditions:
- select a spacing layout for fuel assemblies in the protective container;
- develop a calculation model;
- calculate temperature fields in the storage facility.
2. Preparion of samples from barrier materials for investigation:
- cut out BN-350 fuel assemblies;
- transport irradiated materials from Aktau-city (BN-350) to Kurchatov-city (IAE);
- prepare samples taken out from fuel assembly ducts, protective cask and casing materials for study.
1. Thermal aging of the irradiated material, protective cask and casing samples at the temperature range of 200700С from several minutes to 1 (or 2) years.
2. Study irradiated and non-irradiated materials after their thermal aging in order to determine a corrosion level:
- optical metallography;
- electron microscopy;
- spectrometry (elemental analysis);
- X-ray diffraction phase analysis;
- density measurement.
3. Conduction of corrosion and mechanical tests of the materials after their thermal aging.
4. Theoretical analysis of the experimental results:
- study the possibilities of Arrhenius extrapolation to predict material properties under dry storage;
- study the corrosion process to correctly determine an activation energy value and define factors associated to changes of the material properties during aging;
- simulate the 10-50 years low-temerature aging of the barrier materials using experimental data on their accelerated corrosion at high temperatures.
5. Prediction of BN-350 fuel assembly materials degradation level under the long-term dry storage.
6. Analysis of an environmental risk posed by the dry storage facility as a result of protective barriers degradation after the 50 years storage of the BN-350 fuel assemblies.

Samples for research will be cut out from dusts of the BN-30 spent fuel assemblies made from 12X18H10T or 08X16H11M3 austenitic stainless steel that were irradiated up to 84 dpa at the temperature of 280С and higher. With few assumptions these may be considered as a prototype of fuel assembly cladding – material of the first protective barrier precluding release of radionuclides into the environment. Non-irradiated materials of the protective cask and casing (stainless and carbon steel) accordingly are also planned for investigation. Thermal aging of irradiated and non-irradiated samples is planned to be conducted at 200700С within several minutes and 1 (or 2) years in vacuum, inert gas and air mediums. A corrosion level will be determined by studying the surface layer of materials using optical metallography, scanning and transmission electron microscopy with element analysis capabilities, X-ray diffraction analysis and density measurements. Corrosion tests are planned to be conducted by various methods including ASTM A262 and/or electrochemical potentiodynamic reactivation method. Mechanical tests will be mostly conducted by bending the samples with deformation rates of 10-410-1 с-1 at room temperature. The material density will be defined by hydrostatic weighing and picknometer method. As a result of this work, experimental data will be obtained on corrosion level of the materials depending on temperature and duration of thermal aging, and kinetics of corrosion and mechanical property changes will be evaluated. When the activation energy value of corrosion process is determined using experimental results including the accelerated corrosion data at high temperatures (500-700С), the 10-50 years thermal aging of barrier material will be simulated. To provide this, the Arrhenius extrapolation method is planned to be used introducing experimentally obtained factors that take into account changes in the material corrosion properties during aging process. Based on the obtained results, a prediction will be made for condition of the BN-350 spent fuel assemblies after their 50-year dry storage. A level of the environmental risk posed by possible release of radionuclides into the environment will be estimated upon results of prediction data of barrier material degradation.

As a result of the project implementation, data related to the following are expected:

· temperature fields in fuel assembly under dry storage conditions;
· changes in microstructure, density, corrosion and mechanical properties of the irradiated and non-irradiated materials under long-term thermal aging;
· 10-50 years thermal aging simulation of the irradiated and non-irradiated materials on the basis of experimental data on the accelerated corrosion at high temperatures of aging;
· BN-350 fuel assemblies and protective cask conditions after dry storage for 50 years;
· level of environmental contamination risk as a result of degradation of protective barrier protective properties under 50-year dry storage of the BN-350 fuel assemblies.

Research results obtained under this project may be useful in developing technologies on handling spent fuel assemblies of the power nuclear reactors under long-term dry storage conditions. These results may also contribute much to the basic science because new knowledge on kinetics of the long-term thermal aging of high dose irradiated materials at lower temperatures is expected. The results of investigation will promote to deeper understanding of diffusion processes, and impurity segregation in solid solution under conditions of simultaneous evolution of radiation defects and helium atoms.

Three organizations of the Republic of Kazakhstan participate in the project: the Institute of Atomic Energy (IAE), the Mangyshlak Atomic Energy Combine (MAEC) and the Institute of Nuclear Physics (INP) that previously conducted similar researches. In particular, a preliminary analysis was carried out by means of computer codes for safe transportation and long-term storage of the BN-350 fuel assemblies. Also, some microstructural studies of the BN-350 core materials were performed. Thus, the research planned under this project is natural continuation of these activities.

The project involves scientists and specialists in the field of thermal physics and studies of radiation damage influence on reactor material microstructure and properties. Numerous articles have been published on the project subject and scientific-and-technical reports were issued. Some of the publications of the project participants directly refer to the long-term thermal aging of irradiated steels. The equipment, devices and means necessary for project implementation are available. Operations in the hot cells, equipment for preparing samples, thermal treatment, conducting corrosion and mechanical tests, precision microstructural studies and density measurement are set.

The project meets the ISTC goals: provides an opportunity for weapons scientists and engineers to reorient their activities into a peaceful activity, promotes to integration of Kazakhstan scientists into the International Scientific Community, supports basic and applied research activities in the field of safe storage of spent nuclear materials, contributes to resolving national and International engineering problems related to dry storage of the nuclear reactor spent fuel assemblies.

The Argon National Laboratory (US) will be a project collaborator. However, the project authors are open for cooperation with other interested organizations and scientists from the US and other countries. Collaborators will participate in technical monitoring of Project activities performed by ISTC staff. The collaborators will coordinate the project activities mainly via email and Internet, and discuss results of the investigations with Kazakhstan scientists during mutual visits.


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.

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