L.C. Walters

Thirty years of fuels and materials information from EBR-II

Abstract The Experimental Breeder Reactor-II (EBR-II) was a 62.5 MWt–20 MWe sodium cooled fast reactor that was operated successfully for 30 years. Over its period of operation a wealth of fuels and materials information originated from EBR-II. Several missions were conducted in EBR-II, all of which yielded new and valuable additions to the worlds knowledge base for nuclear materials. Some of the first pioneering experiments on irradiation effects in stainless steels were conducted in EBR-II. Later, practical manifestations of enhanced irradiation creep, swelling, and loss of ductility were experienced on EBR-II components. In addition, for a period of more than 15 years, the EBR-II reactor would become the primary irradiation facility for all fast reactor fuels and materials research and development. Both the initial mission and final mission for EBR-II (the Integral Fast Reactor Concept, IFR), involved the remote reprocessing and irradiation of fast reactor metallic fuels. The fuels and materials information gleaned from these missions will be summarized with the intent of portraying a sample of the valuable legacy that EBR-II contributed to the worlds store of nuclear fuels and materials knowledge.

Status of LMR fuel development in the United States of America

Three fuel systems oxide, metal, and carbide are shown to be reliable to high burnup and a fourth system, nitride, is shown to have promise for LMR applications. The excellent steady state performance of the oxide and metal driver fuels for FFTF and EBR-II, respectively, supported by the experience base of tens of thousands of test pins is provided. Achieving 300 MWd/kg in the oxide fuel system through the use of low swelling cladding and duct materials and the Integral Fast Reactor (IFR) concept that utilizes metallic fuel are described. Arguments for economic viability are presented. Responses to operational transients and severe over-power events are shown to have large safety margins and run-beyond-claddingbreach (RBCB), is shown to be non-threatening to LMR reactor systems. Results from a joint U.S.-Swiss carbide test that operated successfully at high power and burnup in FFTF are also presented.

Irradiation behavior of metallic fast reactor fuels

Abstract Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMRs). In the late 1960s worldwide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970s the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratorys Experimental Breeder Reactor II. The 1980s spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

In-reactor deformation of solution annealed type 304L stainless steel

Abstract Over the past six years at EBR-II, a great deal of information has been obtained on the in-reactor behaviour of solution annealed-Type 304L stainless steel. This information consists of the following: (1) Irradiation induced swelling results in the form of immersion density and transmission electron microscope (TEM) measurements on unstressed material that extends over a temperature range of 395° to 530°C and a neutron fluence range of 1.8 to 9.3 × 10 22 n/cm 2 ( E > 0.1 MeV). (2) Irradiation induced creep results from helium pressurized capsules irradiated at a temperature of 415°C. The hoop stress range covered in the experiment was 0 to 27.3 ksi, and the peak neutron fluence obtained to date is 7 × 10 22 n/cm 2 ( E > 0.1 MeV). (3) Residual stress measurements (slit tube technique) with complementary TEM gradient studies on stressed and unstressed capsules. (4) Comparative swelling studies of stressed cladding material and unstressed capsule material from encapsulated EBR-II driver fuel experiments over wide ranges of temperature and neutron fluences. The deformation information derived from the four above studies represent an extensive data base from which to obtain an understanding of the in-reactor deformation of austenitic stainless steel. It is the purpose of this paper to review our information on the in-reactor deformation of solution annealed Type 304L stainless steel.

Nuclear fuel considerations for the 21st century

Abstract There are many external influences that may control the path that nuclear power deployment follows. In the next 50 years several events may unfold. Fear of the consequences of the greenhouse effect may produce a carbon tax that would make nuclear power economically superior very quickly. This, in turn, would increase the rate at which uranium reserves diminish due to the increased rate of nuclear power deployment. However, breakthroughs in the extraction of uranium from the sea or deployment of fast breeder reactors would greatly extend the uranium reserves and, as well, utilize the thorium cycle. On the other hand, carbon sequestering technology breakthroughs could keep fossil fuels dominant for the remainder of the century. Nuclear power may only then continue, as today, in a lesser role or even diminish. Fusion power or new developments in solar power could completely displace nuclear power as we know it today. Even more difficult to predict is when the demand for mobile fuel for transportation will develop such that hydrogen and hydrogen rich fuel cells will be in common use. When this happens, nuclear power may be the energy source of choice to produce this fuel from water or methane. In a similar vein, the demand for potable and irrigation water may be another driver for the advent of increased deployment of nuclear power. With all these possibilities of events that could happen it appears impossible to predict with any certainty which path nuclear power deployment may take. However, it is necessary to define a strategy that is flexible enough to insure that when a technology is needed, it is ready to be deployed. For the next few decades there will be an evolutionary improvement in the performance of uranium oxide and mixed uranium oxide-plutonium oxide (MOX) LWR fuels. These improvements will be market driven to keep the cost of fuel and the resulting cost of nuclear power electricity as competitive as possible. The development of fuels for accelerator transmutation and for reactor transmutation with inert matrix fuels is in its infancy. A great deal of research has been initiated in a number of countries, which has been summarized in recent conferences. In Europe the work on these fuels is directed at the same problem as their utilization of MOX; namely to reduce the inventory of separated plutonium, minor actinides, and Long Lived Fission Products (LLFP). In the United States there is no reprocessing and thus no inventory of separated civilian plutonium. However, in the United States there is a resistance to a permanent spent fuel repository and thus accelerator transmutation presents a possible alternative. If nuclear power does have a long-term future, then the introduction of the fast reactor is inevitable. Included in the mission of the fast reactors would be the elimination of the inventory of separated plutonium while generating useful energy. The work that is ongoing now on the development of fuel concepts for assemblies that contain actinides and LLFP would be useful for fast reactor transmutation. There is still a great deal of work required to bring the fast breeder reactor option to maturity. Fortunately there is perhaps a fifty-year period to accomplish this work before fast breeders are necessary. With regard to fast reactor fuel development, future work should be considered in three stages. First, all the information obtained over the past forty years of fast reactor fuel development should be completely documented in a manner that future generations can readily retrieve and utilize the information. Fast reactor development came to such an abrupt halt world-wide that a great deal of information is in danger of being lost because most of the researchers and facilities are rapidly disappearing. Secondly, for all of the existing fast reactor fuels, and this includes, oxides, carbides, nitrides, and metallic fuels, the evolutionary work was far from being completed. Although mixed oxide fuels were probably the furthest advanced, there were many concepts for improved claddings and advanced fabrication methods that were never fully explored. Finally, with such an extended period before fast reactors are needed there is ample time for truly innovative fuels to be developed that are capable of performing over a wide range of conditions and coolants.

Neutron irradiation-induced creep of helium pressurized 304L stainless steel capsules

Abstract Irradiation-induced creep and swelling have been measured on 1.5 m long pressurized capsules of solution annealed type 304L stainless steel at 385 °C to neutron doses of 45 dpa. The core-midplane results (fixed position) which have a constant average neutron energy and dose rate but varying time are compared to data taken along the length of the capsule which have constant time but varying average neutron energy and dose rates. Additionally, the effect of stress on swelling, the stress dependency of in-reactor creep and the correlation of irradiation-induced creep and swelling are analyzed utilizing the data generated in this experiment. The results of these analyses are then used as a basis for appraising current theories on irradiation creep.

The relationship between carbide precipitation and the in-reactor deformation of type 316 stainless steel

Abstract The precipitation of carbides from type 316 stainless steel is associated with enhanced irradiation-induced creep and swelling. Carbide precipitation occurs in irradiated type 316 materials at lower temperatures than in their unirradiated counterparts. Once the precipitate is formed, there is little tendency for it to dissolve back into the matrix upon further irradiation; thus, an irreversible change in the materials response to neutron irradiation occurs. The degree of precipitation and its effects on in-reactor strain have been investigated, and the results were used to explain double peaks in diameter profiles of irradiated fuel elements.

Neutron irradiation and compatibility testing of Li2O

A study was made of the neutron irradiation behavior of 6Li-enriched Li2O in EBR-II. In addition, a stress corrosion study was performed ex-reactor to test the compatibility of Li2O with a variety of stainless steels. The irradiation tests showed that tritium and helium retention in the Li2O (∼ 89% dense) lessened with neutron exposure, and the retentions appear to approach a steady-state after ∼ 1% 6Li burnup. The stress corrosion studies, using 316 stainless steel (Ti-modified) and a 35% Ni alloy, showed that stress does not enhance the corrosion, and that dry Li2O is not significantly corrosive, the LiOH content producing the corrosive effects. Corrosion, in general, was not severe because a passivation in sealed capsules seemed to occur after a time which greatly reduced corrosion rates.

In-reactor stress relaxation of Inconel X75O springs

Abstract In 1965 eight surveillance subassemblies were placed in row 12 of the EBR-II sodium-cooled fast breeder reactor with an irradiation temperature near the sodium-inlet temperature of 371°C. At the same time, two other surveillance subassemblies were placed in the primary storage basket, which receives minimal neutron exposure but is immersed in primary sodium and experiences a temperature of 371°C. Each of the subassemblies contained 18 preloaded springs made of Inconel X750. Springs from four of the in-core subassemblies and one subassembly from the storage basket have been evaluated to determine irradiation-enhanced deformation rates to neutron exposures of 4.2 dpa. It was found that the creep coefficient derived from the stress relaxation measurements on Inconel X750 springs was 1.0 × 10 −12 (Pa-dpa) −1 for springs irradiated up to 4.2 dpa (3751 d) at an in-reactor temperature of 371°C. The relaxation behavior was adequately described by a creep law that was linear in neutron fluence and applied stress. Springs encapsulated in helium showed identical in-reactor relaxation rates to springs exposed to the flowing primary sodium. The creep coefficient derived from the present work on Inconel X750 springs was shown to be the same as the creep coefficients determined from various austenitic stainless steel alloys.

The effect of phase instability of stainless steel on the cladding deformation profiles of LMFBR fuel elements

Abstract Enhanced deformation rates in LMFBR fuel element cladding are shown to be consistent with the occurrence of carbide precipitates. Swelling and creep equations, when modified to include the effects of carbide precipitation, account for such phenomena as the second diameter peak and certain large heat-to-heat variations observed in the cladding of irradiated elements.