In situ Investigation of Oxide Films on Zirconium Alloy in PWR Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

Zirconium alloys are used as fuel cladding materials in nuclear power reactors, because these materials have a very low thermal neutron capture cross section as well as desirable mechanical properties. However, the Fukushima accident shows that the oxidation behavior of zirconium alloy is an important issue because the zirconium alloy functions as a shield of nuclear material (i.e., uranium, fission gas), and the degradation on zirconium cladding directly causes severe accident on nuclear power plant. Therefore, to ensure the safety of nuclear power reactors, the performance and sustainability of nuclear fuel should be understood. Currently, the water-metal interface is regarded as the rate-controlling site governing the rapid oxidation transition in high-burn-up fuels. Zirconium oxide is formed at the water-metal interface, and its structure and phase play an important role in determining its mechanical properties. In the early stage of the oxidation process, zirconium oxide with both tetragonal and monoclinic phases is formed. With an increase in the oxidation time to 150 h, the unstable tetragonal phase disappears and the monoclinic phase is dominant and possibly because of the stress relaxation according to previous and present results.

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

2018-03-31

Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

Development of zirconium alloy tube manufacturing technology

International Nuclear Information System (INIS)

Kim, In Kyu; Park, Chan Hyun; Lee, Seung Hwan; Chung, Sun Kyo

2009-01-01

In late 2004, Korea Nuclear Fuel Company (KNF) launched a government funded joint development program with Westinghouse Electric Co. (WEC) to establish zirconium alloy tube manufacturing technology in Korea. Through this program, KNF and WEC have developed a state of the art facility to manufacture high quality nuclear tubes. KNF performed equipment qualification tests for each manufacturing machine with the support of WEC, and independently carried out product qualification tests for each tube product to be commercially produced. Apart from those tests, characterization test program consisting of specification test and characterization test was developed by KNF and WEC to demonstrate to customers of KNF the quality equivalency of products manufactured by KNF and WEC plants respectively. As part of establishment of performance evaluation technology for zirconium alloy tube in Korea, KNF carried out analyses of materials produced for the characterization test program using the most advanced techniques. Thanks to the accomplishment of the development of zirconium alloy tube manufacturing technology, KNF is expected to acquire positive spin off benefits in terms of technology and economy in the near future

Performance of U-Pu-Zr fuel cast into zirconium molds

International Nuclear Information System (INIS)

Crawford, D.C.; Lahm, C.E.; Tsai, H.

1992-01-01

Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium--2.5 wt percent niobium and zirconium--1.1 wt percent chromium--0.1 wt percent iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium--niobium and zirconium--chromium--iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 percent a cyclic frequency exceeding 0.116 Hz (10,000 cycles/ day) would be required to cause fatigue failure of the sheath before its design life is realized

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium-2.5 wt% niobium and zirconium-1.1 wt% chronium-0.1 wt% iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium-niobium and zirconium-chromium-iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 per cent a cyclic frequency exceeding 0.116 Hz (10 000 cycles/day) would be required to cause fatigue failure of the sheath before its design life is realized. (author)

Development of microstructure in thermomechanical processing of zirconium alloys

International Nuclear Information System (INIS)

Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

Thermofluency in zirconium alloys

International Nuclear Information System (INIS)

Orozco M, E.A.

1976-01-01

A summary is presented about the theoretical and experimental results obtained at present in thermofluency under radiation in zirconium alloys. The phenomenon of thermofluency is presented in a general form, underlining the thermofluency at high temperature because this phenomenon is similar to the thermofluency under radiation, which ocurrs in zirconium alloys into the operating reactor. (author)

Surface coating Zr or Zr alloy nuclear fuel elements

International Nuclear Information System (INIS)

Donaghy, R.E.; Sherman, A.H.

1980-01-01

A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Zirconium behaviour during electrorefining of actinide-zirconium alloy in molten LiCl-KCl on aluminium cathodes

Energy Technology Data Exchange (ETDEWEB)

Meier, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany); Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Malmbeck, R.; Krachler, M.; Rodrigues, A.; Claux, B.; Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Fanghänel, Th. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany)

2016-04-15

A pyrochemical electrorefining process for the recovery of actinides from metallic nuclear fuel based on actinide-zirconium alloys (An–Zr) in a molten salt is being investigated. In this process actinides are group-selectively recovered on solid aluminium cathodes as An–Al alloys using a LiCl–KCl eutectic melt at a temperature of 450 °C. In the present study the electrochemical behaviour of zirconium during electrorefining was investigated. The maximum amount of actinides that can be oxidised without anodic co-dissolution of zirconium was determined at a selected constant cathodic current density. The experiment consisted of three steps to assess the different stages of the electrorefining process, each of which employing a fresh aluminium cathode. The results indicate that almost a complete dissolution of the actinides without co-dissolution of zirconium is possible under the applied experimental conditions. - Highlights: • Recovery of actinides was shown by electrorefining of U/Pu–Zr alloys in LiCl–KCl. • Constant current density of 20 mA/cm{sup 2} is applied. • Most of the actinides were dissolved avoiding zirconium co-dissolution. • Deterioration of the deposit quality by a small amount of co-deposited Zr is not observed.

Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

International Nuclear Information System (INIS)

Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

2004-01-01

Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows f or surface

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

International Nuclear Information System (INIS)

Rosinger, H.E.; Northwood, D.O.

1979-01-01

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

International Nuclear Information System (INIS)

Onimus, Fabien

2015-01-01

Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

METMET fuel with Zirconium matrix alloys

International Nuclear Information System (INIS)

Savchenko, A.; Konovalov, I.; Totev, T.

2008-01-01

The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

Applications for zirconium and columbium alloys

International Nuclear Information System (INIS)

Condliff, A.F.

1986-01-01

Currently, zirconium and columbium are used in a wide range of applications, overlapping only in the field of corrosion control. As a construction material, zirconium is primarily used by the nuclear power industry. The use of zirconium in the chemical processing industry (CPI) is, however, increasing steadily. Columbian alloys are primarily applied as superconducting alloys for research particle accelerators and fusion generators as well as in magnetic resonance imaging for medical diagnosis

Modelling of zirconium alloys corrosion in LWRs

International Nuclear Information System (INIS)

Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

1999-01-01

Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

Status and task of the study on the hydrogen embrittlement of zirconium alloys

International Nuclear Information System (INIS)

Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

1995-08-01

As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

Sliding wear and friction behavior of zirconium alloy with heat-treated Inconel718

Energy Technology Data Exchange (ETDEWEB)

Kim, J.H., E-mail: kimjhoon@cnu.ac.kr [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.M. [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.K.; Jeon, K.L. [Nuclear Fuel Technology Department, Korea Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2014-04-01

In water-cooled nuclear reactors, the sliding of fuel rod can lead to severe wear and it is an important issue to sustain the structural integrity of nuclear reactor. In the present study, sliding wear behavior of zirconium alloy in dry and water environment using Pin-On-Disk sliding wear tester was investigated. Wear resistance of zirconium alloy against heat-treated Inconel718 pin was examined at room temperature. Sliding wear tests were carried out at different sliding distance, axial load and sliding speed based on ASTM (G99-05). The results of these experiments were verified with specific wear rate and coefficient of friction. The micro-mechanisms responsible for wear in zirconium alloy were identified to be microcutting and microcracking in dry environment. Moreover, micropitting and delamination were observed in water environment.

New zirconium alloys for nuclear application; Novas ligas de zirconio para aplicacao nuclear

Energy Technology Data Exchange (ETDEWEB)

Lobo, R.M.; Andrade, A.H.P., E-mail: rmlobo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2010-07-01

Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions ({approx} 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

Artefacts in multimodal imaging of titanium, zirconium and binary titanium-zirconium alloy dental implants: an in vitro study.

Science.gov (United States)

Smeets, Ralf; Schöllchen, Maximilian; Gauer, Tobias; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

2017-02-01

To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium-zirconium alloy dental implants. Zirconium, titanium and titanium-zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line-distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. While titanium and titanium-zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium-zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium-zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium-zirconium alloy induced more severe artefacts than zirconium and titanium. MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium-zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting.

Artefacts in multimodal imaging of titanium, zirconium and binary titanium–zirconium alloy dental implants: an in vitro study

Science.gov (United States)

Schöllchen, Maximilian; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

2017-01-01

Objectives: To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium–zirconium alloy dental implants. Methods: Zirconium, titanium and titanium–zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line–distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. Results: While titanium and titanium–zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium–zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium–zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium–zirconium alloy induced more severe artefacts than zirconium and titanium. Conclusions: MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium–zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting. PMID:27910719

Zirconium - an imported mineral commodity

International Nuclear Information System (INIS)

1983-10-01

This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

International Nuclear Information System (INIS)

Verlet, Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

Study of corrosion kinetics of fuel element tubes from calcium-thermal zirconium alloy Zr1Nb in water at 350 degree C and in vapour at 400 and 500 degree C

International Nuclear Information System (INIS)

Petel'guzov, I.A.

2002-01-01

In the report brought results of corrosion process studies in water medium of pipe samples for fuel element shells from Zr1Nb alloy (earlier KTZ-110),made from the calcium-thermal zirconium alloys developed in the Ukraine of technology and,for the comparison,samples of pipes from the staff alloy E110, applicable in fuel elements acting reactors of type WWER. Tests were conducted under the working temperature of fuel shells in the reactor (350 degree C) in during of 14000 hours and under increased temperatures (400 degree C) within a time acordinly 4000 hours. Samples from the alloy Zr1Nb had more high contents of oxygen (before 0,12%...0,16%), than staff alloy Eh110 (0,08%O). Studies have shown sufficiently high corrosion stability of experimental alloy Zr1Nb, close to stability of alloy E110.Discovered signs of corrosion 'breakway' or 'transition' on kinetic corrosion curves of Zr1Nb alloys and E110 alloy, characterisating zircaloy type of alloy. Considered mechanism of influence of oxygen on the corrosion process of zirconium alloys with the additive a niobium

Application of FEM analytical method for hydrogen migration behaviour in Zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Arioka, K; Ohta, H [Takasago Research and Development Center, Mitsubishi Heavy Industries Ltd, Hyogo-ken (Japan)

1997-02-01

It is well recognized that the hydriding behaviours of Zirconium alloys are very significant problems as a safety issues. Also, it is well known that the diffusion of hydrogen in Zirconium alloys are affected not only by concentration but also temperature gradient. But in actual component, especially heat transfer tube such as fuel rod, we can not avoid the temperature gradient in some degree. So, it is very useful to develop the computer code which can analyze the hydrogen diffusion and precipitation behaviours under temperature gradient as a function of the structure of fuel rod. For this objective, we have developed the computer code for hydrogen migration behaviour using FEM analytical methods. So, following items are presented and discussed. Analytical method and conditions; correlation between the computed and test results; application to designing studies. (author). 8 refs, 4 figs, 2 tabs.

Environmentally-induced cracking of zirconium alloys - a review

International Nuclear Information System (INIS)

Cox, B.

1990-01-01

The general field of environmentally-induced cracking of zirconium alloys has been reviewed and the phenomena that are observed and the progress in understanding the mechanisms are summarized. The details of the industrially important pellet-clad interaction failures of nuclear reactor fuel have been left for a companion review, and only observations on the mechanism are summarized briefly here. It is concluded that in the zirconium alloy system, by virtue of the physical peculiarities of the system, it is easier to reach unambiguous conclusions about the environmental cracking mechanisms that are operating than with other systems. Thus, chemical dissolution in either liquid or vapour phase is thought to be the principal mechanism for intergranular cracking, while adsorption-induced embrittlement is thought to be the most common transgranular quasi-cleavage process. Hydrogen embrittlement in this system can be identified because it requires precipitated hydride that gives characteristic fractography when cracked. Only in a few instances does stress-corrosion cracking appear to proceed by a hydride cracking mechanism. (orig.)

Modification of structural phase state in superficial layers of fuel tubes made of Zirconium alloys

International Nuclear Information System (INIS)

Volkov, N.; Kalin, B.; Pimenov, Y.; Timoshin, S.

2011-01-01

The paper presents the results obtained in developing the method for introduction of the required changes into states and properties of outer surface on fuel rod cladding made of zirconium alloys E110 and E635 through irradiation by radial Ar + ion beam with a broad energy spectrum. In particular, the paper demonstrates that ion beam treatment of the claddings surface, at the final stage of their fabrication, can upgrade substantially quality of outer tubular surface after mechanical polishing (the cleaner surface, the lower roughness, removal of technological transversal scratches). In addition, the ion beam irradiation results in higher micro-hardness of the modified layer and in better tribological parameters. Kinetic effects in growth of oxide films were studied for the tubular samples of zirconium alloys after ion-beam treatment (cleaning and polishing by radial Ar + ion beam). Also, corrosion tests of the tubular samples were carried out in water (at 350 0 C) and steam (at 350, 375 and 400 0 C) with duration up to 3000 hours. It was revealed that oxide layer consisting mainly of zirconium dioxide in monoclinic modification was formed on tubular surface after oxidation at 3500 0 C in water or steam. The oxidizing process in the pressurized steam created thicker oxide layer on tubular surface than that in the pressurized water. Experimental data were used to determine optimal conditions for ion-beam treatment of outer fuel tube surface. The tubular samples with the following geometrical parameters were investigated: length - up to 500 mm, diameter - 9,15 mm. Optimal regimes for ion-beam cleaning and polishing of the tubular samples were studied up to the process rate of 1 meter per minute. Within the frames of linear approximation, analytical relationships were derived for time dependent growth of oxide films and used to evaluate thickness of oxide film under test conditions (duration . up to 10000 hours). Thickness of oxide films can cover the range from 6 �

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

Science.gov (United States)

Shor, R.S.; Vogler, S.

1958-01-21

A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

Advances in zirconium technology for nuclear reactor application

International Nuclear Information System (INIS)

Ganguly, C.

2002-01-01

Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

Delayed hydride cracking of zirconium alloy fuel cladding

International Nuclear Information System (INIS)

2010-10-01

This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

Protection of zirconium and its alloys by metallic coatings

International Nuclear Information System (INIS)

Loriers, H.; Lafon, A.; Darras, R.; Baque, P.

1968-01-01

At 600 deg. C in an atmosphere of carbon dioxide, zirconium and its alloys undergo corrosion which presents two aspects simultaneously: - formation of a surface layer of zirconia, - dissolution of oxygen in the alloy sub-layer leading to brittleness. The two phenomena greatly restrict the possibilities of using zirconium alloys as a canning material for fuel elements in CO 2 cooled nuclear reactors. An attempt has thus been made to limit, and perhaps to suppress, the corrosion effects in zirconium under these conditions by protecting it with metallic coatings. A first attempt to obtain a protection using copper-based coatings did not produce the result hoped for. Aluminium coatings produced by vacuum evaporation, followed by a consolidating thermal treatment make it possible to prevent the formation of the zirconia layer, but they do not eliminate the hardening effect produced by oxygen diffusion. On the other hand, electrolytically produced chromium deposits whose adherence is improved by a thermal vacuum treatment, counteract both these phenomena simultaneously. A similar result has been obtained with coatings of molybdenum produced by the technique of high-frequency inductive plasma sputtering. The particular effectiveness of the last two types of coatings is due to their structures characterized by the existence of an adherent film of chromium or molybdenum in the free state. (authors) [fr

Stainless steel-zirconium alloy waste forms

International Nuclear Information System (INIS)

McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

1996-01-01

An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

Critical assessment of finite element analysis applied to metal–oxide interface roughness in oxidising zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

2015-09-15

As a nuclear fuel cladding material, zirconium alloys act as a barrier between the fuel and pressurised steam or lithiated water environment. Controlling degradation mechanisms such as oxidation is essential to extending the in-service lifetime of the fuel. At temperatures of ∼360 °C zirconium alloys are known to exhibit cyclical, approximately cubic corrosion kinetics. With acceleration in the oxidation kinetics occurring every ∼2 μm of oxide growth, and being associated with the formation of a network of lateral cracks. Finite element analysis has been used previously to explain the lateral crack formation by the development of localised out-of-plane tensile stresses at the metal–oxide interface. This work uses the Abaqus finite element code to assess critically current approaches to representing the oxidation of zirconium alloys, with relation to undulations at the metal–oxide interface and localised stress generation. This includes comparison of axisymmetric and 3D quartered modelling approaches, and investigates the effect of interface geometry and plasticity in the metal substrate. Particular focus is placed on the application of the anisotropic strain tensor used to represent the oxidation mechanism, which is typically applied with a fixed coordinate system. Assessment of the impact of the tensor showed that 99% of the localised tensile stresses originated from the out-of-plane component of the strain tensor, rather than the in-plane expansion as was previously thought. Discussion is given to the difficulties associated with this modelling approach and the requirements for future simulations of the oxidation of zirconium alloys.

A microstuctural study on accelerated zirconium alloy oxidation

International Nuclear Information System (INIS)

Sohn, Seung Bum; Oh, Seung Jun; Jang, Jung Nam; Kim, Yong Soo; Jung, Yong Hwan; Baek, Jong Hyuk; Park, Jung Yong

2005-01-01

It has been reported that the effect of thermal redistribution of hydrides across the zirconium metaloxide interface, coupled with thermal feedback on the metal-oxide interface, is a dominating factor in the accelerated oxidation in zirconium alloys cladding PWR fuel. Basically this influence determines characteristic of oxide layer. Influence estimation for corrosion oxide layer due to hydrogen / hydride carried out because of investigation on the kinetic on accelerated oxidation due to hydride precipitation was preceded. Generally, it is known that ZrO 2 tetragonal layer structures play an important role as a barrier layer. So analysing the ZrO 2 monoclinic and tetragonal structure distribution is our main aim. Especially, this study focused on the hydride effects. In other words, the difference of crystal structure distribution between pre-hydrided and without hydrided specimen is just expected results. Experimental results of microstructure at zirconium metal-oxide interface through TEM and EBSD analysis was confirmed

Development of new zirconium alloys for PWR fuel rod claddings

International Nuclear Information System (INIS)

Zhao Wenjin; Zhou Bangxin; Miao Zhi; Li Cong; Jiang Hongman; Yu Xiaowei; Jiang Yourong; Huang Qiang; Gou Yuan; Huang Decheng

2001-01-01

An advanced zirconium alloys containing Sn, Nb, Fe and Cr have been developed. The relationships between manufacturing, microstructure and corrosion performance for the new alloys have been studied. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in lithia water at 633 K and high-temperature steam at 773 K. Analytical electron microscopy demonstrated that the best out-of-pile corrosion performance was obtained for microstructure containing a fine and uniform distribution of β-Nb and Zr(Fe, Nb) 2 particles. Autoclave testing in LiOH solution indicated that two kinds of alloys (N18, N36) showed the lower corrosion rate than the reference Zr-4 tested, and especially, the corrosion resistance in superheated steam at 773 K was much better. Moreover, the mechanical properties were superior to Zr-4. And the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the oxide thickness

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

Jeong, Yong Hwan; Baek, B. J.; Park, S. Y. and others

1999-08-01

The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs

Superconductivity in zirconium-rhodium alloys

Science.gov (United States)

Zegler, S. T.

1969-01-01

Metallographic studies and transition temperature measurements were made with isothermally annealed and water-quenched zirconium-rhodium alloys. The results clarify both the solid-state phase relations at the Zr-rich end of the Zr-Rh alloy system and the influence upon the superconducting transition temperature of structure and composition.

Fuel assembly and fuel cladding tube

International Nuclear Information System (INIS)

Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

1996-01-01

A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

A half-century of changes in zirconium alloys

International Nuclear Information System (INIS)

Mardon, J.P.; Barberis, P.; Hoffmann, P.B.

2008-01-01

This article presents the history of zirconium alloys for PWR and BWR technologies. For more than 20 years zirconium alloys have evolved to cope with demands of the reactor operators concerning the burn-up extension and new safety margins. The poor properties of Zircaloy-1 concerning corrosion have led researchers to add elements like iron by developing Zircaloy-3A and Zircaloy-3C, and resulting in Zircaloy-4 with tin addition (from 1.30% to 1.50%). Zircaloy-4 is now outdated for PWR and new zirconium alloys with niobium are used (M5, ZIRLO...) they present a better resistance to corrosion, to hydridation, to creep and they are less prone to dimensional changes under irradiation. (A.C.)

Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

1975-09-01

This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

Manufacturing process to reduce large grain growth in zirconium alloys

International Nuclear Information System (INIS)

Rosecrans, P.M.

1987-01-01

A method is described of treating cold worked zirconium alloys to reduce large grain growth during thermal treatment above its recrystallization temperature. The method comprises heating the zirconium alloy at a temperature of about 1300 0 F. to 1350 0 F. for about 1 to 3 hours subsequent to cold working the zirconium alloy and prior to the thermal treatment at a temperature of between 1450 0 -1550 0 F., the thermal treatment temperature being above the recrystallization temperature

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

Energy Technology Data Exchange (ETDEWEB)

Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

ZIRCONIUM-TITANIUM-BERYLLIUM BRAZING ALLOY

Science.gov (United States)

Gilliland, R.G.; Patriarca, P.; Slaughter, G.M.; Williams, L.C.

1962-06-12

A new and improved ternary alloy is described which is of particular utility in braze-bonding parts made of a refractory metal selected from Group IV, V, and VI of the periodic table and alloys containing said metal as a predominating alloying ingredient. The brazing alloy contains, by weight, 40 to 50 per cent zirconium, 40 to 50 per cent titanium, and the balance beryllium in amounts ranging from 1 to 20 per cent, said alloy having a melting point in the range 950 to 1400 deg C. (AEC)

Unirradiated UO2 in irradiated zirconium alloy sheathing

International Nuclear Information System (INIS)

MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

1979-07-01

Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

Science.gov (United States)

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Hydrogen content in titanium and a titanium–zirconium alloy after acid etching

Energy Technology Data Exchange (ETDEWEB)

Frank, Matthias J.; Walter, Martin S. [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Lyngstadaas, S. Petter [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Wintermantel, Erich [Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Haugen, Håvard J., E-mail: h.j.haugen@odont.uio.no [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway)

2013-04-01

Dental implant alloys made from titanium and zirconium are known for their high mechanical strength, fracture toughness and corrosion resistance in comparison with commercially pure titanium. The aim of the study was to investigate possible differences in the surface chemistry and/or surface topography of titanium and titanium–zirconium surfaces after sand blasting and acid etching. The two surfaces were compared by X-ray photoelectron spectroscopy, secondary ion mass spectroscopy, scanning electron microscopy and profilometry. The 1.9 times greater surface hydrogen concentration of titanium zirconium compared to titanium was found to be the major difference between the two materials. Zirconium appeared to enhance hydride formation on titanium alloys when etched in acid. Surface topography revealed significant differences on the micro and nanoscale. Surface roughness was increased significantly (p < 0.01) on the titanium–zirconium alloy. High-resolution images showed nanostructures only present on titanium zirconium. - Highlights: ► TiZr alloy showed increased hydrogen levels over Ti. ► The alloying element Zr appeared to catalyze hydrogen absorption in Ti. ► Surface roughness was significantly increased for the TiZr alloy over Ti. ► TiZr alloy revealed nanostructures not observed for Ti.

Techniques for chemical characterization of zirconium and its alloys

International Nuclear Information System (INIS)

Iyer, K.V.; Bassan, M.K.T.; Sudersanan, M.

2002-01-01

Chemical characterization of zirconium and its alloys such as zircaloy, Zr-Nb, etc for minor and trace constituents like Nb, Ti, Fe, Cr, Ni, Sn, Al etc has been carried out. Zirconium, being a major constituent, has been determined by gravimetry as zirconium oxide while other constituents like Nb, Ti, Fe have been determined by spectrophotometric methods. Other metals of importance at trace level have been estimated by AAS or ICPAES. The judicious use of both conventional and modern instrumental methods of analysis helps in the characterization of zirconium and its alloys for various major and minor constituents. The role of matrix effect in the determination was also investigated and methods have been worked out based on a preliminary separation of zirconium by a hydroxide precipitation. (author)

Development of tantalum–zirconium alloy for hydrogen purification

Energy Technology Data Exchange (ETDEWEB)

Kumar, Sanjay, E-mail: sanjay.barc@gmail.com [Fusion Reactor Materials Section, MG, BARC, Mumbai 85 (India); IAMR, Hiroshima University, Higashihiroshima 739-8530 (Japan); Singh, Anamika [GSASM Hiroshima University, Higashihiroshima 739-8530 (Japan); Jain, Uttam; Dey, Gautam Kumar [Fusion Reactor Materials Section, MG, BARC, Mumbai 85 (India)

2016-11-01

Highlights: • Terminal solid solubility of Ta increases with Zr addition. • Increase in lattice parameters of Ta due to Zr addition may be the possible reason. • Enhance H solubility could also be explained on the change in e-DOS of Ta–Zr alloys. • Ta–Zr alloys could be possible combination for hydrogen purification membrane. - Abstract: Terminal solid solubility of hydrogen in Ta–Zr alloys has been studied in connection with the development of tantalum based metallic membrane for hydrogen/tritium purification. The alloys were prepared by vacuum arc melting technique and subsequently cold rolled to 0.2 mm thickness. The terminal solid solubility of hydrogen in these cold rolled samples was investigated in a modified Sieverts apparatus. The terminal solid solubility of hydrogen was marginally increased with zirconium content. The change in the lattices parameter of tantalum upon zirconium addition and the higher affinity of zirconium for hydrogen as compared to tantalum could be the possible reasons.

Study of the processes for of remelting zirconium alloys in an electric arc furnace

Energy Technology Data Exchange (ETDEWEB)

Pereira, Luiz A.T.; Rossi, Jesualdo L.; Costa, Guilherme R.; Martinez, Luis G.; Sato, Ivone M., E-mail: luiz.atp@uol.com.br, E-mail: jelrossi@ipen.br, E-mail: guilhermeramoscosta@gmail.com, E-mail: lgallego@ipen.br, E-mail: imsato@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Zirconium alloy tubes are used as cladding for fuel elements of PWR nuclear reactors, which contains the UO{sub 2} pellets. In the manufacture of these fuel element parts, machining chips from the nuclear grade zirconium alloys are generated. Hence, these machining chips cannot be discarded, as ordinary metallic waste. Thus, the recycling of this material is a strategic aspect for the nuclear technology, both for economic and environmental issues. The main reason is that nuclear grade alloys have very high cost, are not commercially produced in Brazil and has to be imported for the manufacture of the nuclear fuels. This work discusses a method to melt and recycle Zircaloy chips, using an electric-arc furnace to obtain small laboratory ingots. The chemical composition of the ingots was determined using X-ray fluorescence spectroscopy and was compared to the specifications of nuclear grade Zircaloy and to the chemical composition of the received machining chips. The ingots were annealed in high vacuum, as well as were hot rolled in a mill. The microstructures were characterized by optical microscopy. The hardness was evaluated using the Rockwell B scale hardness. The results showed that the compositions of the recycled Zircaloy comply with the chemical specifications and a suitable microstructure has been obtained for nuclear use. (author)

In-situ electrochemical impedance spectroscopy measurements of zirconium alloy oxide conductivity: Relationship to hydrogen pickup

International Nuclear Information System (INIS)

Couet, Adrien; Motta, Arthur T.; Ambard, Antoine; Livigni, Didier

2017-01-01

Highlights: • In-situ electrochemistry on zirconium alloys in 360 °C pure water show oxide layer resistivity changes during corrosion. • A linear relationship is observed between oxide resistivity and instantaneous hydrogen pickup fraction. • The resistivity of the oxide layer formed on Zircaloy-4 (and thus its hydrogen pickup fraction) is higher than on Zr-2.5Nb. - Abstract: Hydrogen pickup during nuclear fuel cladding corrosion is a critical life-limiting degradation mechanism for nuclear fuel. Following a program dedicated to zirconium alloys, corrosion, it has been hypothesized that oxide electronic resistivity determines hydrogen pickup. In-situ electrochemical impedance spectroscopy experiments were performed on Zircaloy-4 and Zr-2.5Nb alloys in 360 °C water. The oxide resistivity was measured as function of time. The results show that as the oxide resistivity increases so does the hydrogen pickup fraction. The resistivity of the oxide layer formed on Zircaloy-4 is higher than on Zr-2.5Nb, resulting in a higher hydrogen pickup fraction of Zircaloy-4, compared to Zr-2.5Nb.

Neutron activation of chlorine in zirconium and zirconium alloys use of the matrix as comparator

International Nuclear Information System (INIS)

Cohen, I.M.; Gomez, C.D.; Mila, M.I.

1981-01-01

A procedure is described for neutron activation analysis of chlorine in zirconium and zirconium alloys. Calculation of chlorine concentration is performed relative to zirconium concentration in the matrix in order to minimize effects of differences in irradiation and counting geometry. Principles of the method and the results obtained are discussed. (author)

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Energy Technology Data Exchange (ETDEWEB)

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov [Pacific Northwest National Laboratory (United States); Tomé, Carlos, E-mail: tome@lanl.gov [Los Alamos National Laboratory (United States); Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com [ANATECH Corporation (United States); Alankar, Alankar, E-mail: alankar.alankar@iitb.ac.in [Indian Institute of Technology Bombay (India); Subramanian, Gopinath, E-mail: gopinath.subramanian@usm.edu [University of Southern Mississippi (United States); Stanek, Christopher, E-mail: stanek@lanl.gov [Los Alamos National Laboratory (United States)

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Design and development of novel MRI compatible zirconium- ruthenium alloys with ultralow magnetic susceptibility.

Science.gov (United States)

Li, H F; Zhou, F Y; Li, L; Zheng, Y F

2016-04-19

In the present study, novel MRI compatible zirconium-ruthenium alloys with ultralow magnetic susceptibility were developed for biomedical and therapeutic devices under MRI diagnostics environments. The results demonstrated that alloying with ruthenium into pure zirconium would significantly increase the strength and hardness properties. The corrosion resistance of zirconium-ruthenium alloys increased significantly. High cell viability could be found and healthy cell morphology observed when culturing MG 63 osteoblast-like cells and L-929 fibroblast cells with zirconium-ruthenium alloys, whereas the hemolysis rates of zirconium-ruthenium alloys are zirconium-ruthenium alloys (1.25 × 10(-6) cm(3)·g(-1)-1.29 × 10(-6) cm(3)·g(-1) for zirconium-ruthenium alloys) are ultralow, about one-third that of Ti-based alloys (Ti-6Al-4V, ~3.5 × 10(-6) cm(3)·g(-1), CP Ti and Ti-6Al-7Nb, ~3.0 × 10(-6) cm(3)·g(-1)), and one-sixth that of Co-Cr alloys (Co-Cr-Mo, ~7.7 × 10(-6) cm(3)·g(-1)). Among the Zr-Ru alloy series, Zr-1Ru demonstrates enhanced mechanical properties, excellent corrosion resistance and cell viability with lowest magnetic susceptibility, and thus is the optimal Zr-Ru alloy system as therapeutic devices under MRI diagnostics environments.

Design and development of novel MRI compatible zirconium- ruthenium alloys with ultralow magnetic susceptibility

Science.gov (United States)

Li, H.F.; Zhou, F.Y.; Li, L.; Zheng, Y.F.

2016-01-01

In the present study, novel MRI compatible zirconium-ruthenium alloys with ultralow magnetic susceptibility were developed for biomedical and therapeutic devices under MRI diagnostics environments. The results demonstrated that alloying with ruthenium into pure zirconium would significantly increase the strength and hardness properties. The corrosion resistance of zirconium-ruthenium alloys increased significantly. High cell viability could be found and healthy cell morphology observed when culturing MG 63 osteoblast-like cells and L-929 fibroblast cells with zirconium-ruthenium alloys, whereas the hemolysis rates of zirconium-ruthenium alloys are alloys and Ti-based alloys, the magnetic susceptibilities of the zirconium-ruthenium alloys (1.25 × 10−6 cm3·g−1–1.29 × 10−6 cm3·g−1 for zirconium-ruthenium alloys) are ultralow, about one-third that of Ti-based alloys (Ti–6Al–4V, ~3.5 × 10−6 cm3·g−1, CP Ti and Ti–6Al–7Nb, ~3.0 × 10−6 cm3·g−1), and one-sixth that of Co–Cr alloys (Co–Cr–Mo, ~7.7 × 10−6 cm3·g−1). Among the Zr–Ru alloy series, Zr–1Ru demonstrates enhanced mechanical properties, excellent corrosion resistance and cell viability with lowest magnetic susceptibility, and thus is the optimal Zr–Ru alloy system as therapeutic devices under MRI diagnostics environments. PMID:27090955

Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

International Nuclear Information System (INIS)

Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

2008-01-01

As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

Research on development and application of titanium and zirconium alloys

International Nuclear Information System (INIS)

Suzuki, Toshiyuki; Sasano, Hisaoki; Uehara, Shigeaki; Nakano, Osamu; Shibata, Michio

1983-01-01

It can be said that titanium and zirconium are new metals from the viewpoint of the history of metals, but both have grown to the materials supporting modern industries, titanium alloys in aerospace and ocean development, and zirconium alloys in nuclear power application. However, the properties of both alloys have not yet been clarified. In this study, the synthesis of TiNi and its properties, precipitation hardening type titanium alloys, and the effect of oxygen on the mechanical properties of both alloys were examined. TiNi is the typical intermetallic compound which shows the peculiar properties. The method of its synthesis by diffusion was examined, and it was clarified that it is useful as a structural material and also as a functional material. Precipitation hardening type alloys have not been developed in titanium alloys, but in this study, the feasibility of several alloy systems was found. Both titanium and zirconium have large affinity to oxygen, and the oxygen absorbed in the manufacturing process cannot be reduced. The tensile property of both alloys was examined in wide temperature range, and the effect of oxygen was clarified. (Kako, I.)

METHOD AND ALLOY FOR BONDING TO ZIRCONIUM

Science.gov (United States)

McCuaig, F.D.; Misch, R.D.

1960-04-19

A brazing alloy can be used for bonding zirconium and its alloys to other metals, ceramics, and cermets, and consists of 6 to 9 wt.% Ni, 6 to 9 wn~.% Cr, Mo, or W, 0 to 7.5 wt.% Fe, and the balance Zr.

Fluorimetric determination of uranium in zirconium and zircaloy alloys

International Nuclear Information System (INIS)

Acosta L, E.

1991-05-01

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

High strength corrosion-resistant zirconium aluminum alloys

International Nuclear Information System (INIS)

Schulson, E.M.; Cameron, D.J.

1976-01-01

A zirconium-aluminum alloy is described possessing superior corrosion resistance and mechanical properties. This alloy, preferably 7.5-9.5 wt% aluminum, is cast, worked in the Zr(Al)-Zr 2 Al region, and annealed to a substantially continuous matrix of Zr 3 Al. (E.C.B.)

Neutronographic Texture Analysis of Zirconium Based Alloys

International Nuclear Information System (INIS)

Kruz'elová, M; Vratislav, S; Kalvoda, L; Dlouhá, M

2012-01-01

Neutron diffraction is a very powerful tool in texture analysis of zirconium based alloys used in nuclear technique. Textures of five samples (two rolled sheets and three tubes) were investigated by using basal pole figures, inversion pole figures, and ODF distribution function. The texture measurement was performed at diffractometer KSN2 on the Laboratory of Neutron Diffraction, Department of Solid State Engineering, Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. Procedures for studying textures with thermal neutrons and procedures for obtaining texture parameters (direct and inverse pole figures, three dimensional orientation distribution function) are also described. Observed data were processed by software packages HEXAL and GSAS. Our results can be summarized as follows: i) All samples of zirconium alloys show the distribution of middle area into two maxima in basal pole figures. This is caused by alloying elements. A characteristic split of the basal pole maxima tilted from the normal direction toward the transverse direction can be observed for all samples, ii) Sheet samples prefer orientation of planes (100) and (110) perpendicular to rolling direction and orientation of planes (002) perpendicular to normal direction, iii) Basal planes of tubes are oriented parallel to tube axis, meanwhile (100) planes are oriented perpendicular to tube axis. Level of resulting texture and maxima position is different for tubes and for sheets. The obtained results are characteristic for zirconium based alloys.

Contribution to the understanding of zirconium alloy deformation under irradiation at high doses

International Nuclear Information System (INIS)

Gharbi, Nesrine

2015-01-01

The growth of zirconium alloy tubes of PWR fuel assemblies is the result of two phenomena: axial irradiation creep and stress 'free' growth which is correlated to the formation of c-loops at high irradiation doses. This PhD work aims at investigating the coupling between these two phenomena through a fine Transmission Electron Microscopy analysis of the effect of a macroscopic applied stress on the c-loop microstructure. 600 keV Zr + ion irradiations were performed at 300 C on two recrystallized zirconium alloys: Zircaloy-4 and M5. Thanks to a device specifically designed, different tensile or compressive stress levels were applied under ion irradiation. The microstructural observations have shown that the c-loop density reduces in grains oriented with the c-axis close to the direction of the applied tensile stress or far from the direction of the applied compressive stress, which is in good agreement with the SIPA mechanism. Nevertheless, the examination of a large number of grains has revealed dispersion from grain to grain. This dispersion, which could be explained by the intergranular heterogeneities, reduces the magnitude of the stress effect on c-loop microstructure. In parallel to this experimental study, a cluster dynamics model has been able to describe the evolution under irradiation of zirconium and Zircaloy-4 microstructure and to assess the effect of stress on c-loop microstructure. On the macroscopic scale, a physical model was also developed to predict the irradiation growth and creep behaviour of zirconium alloy tubes. (author) [fr

Titanium and zirconium alloys

International Nuclear Information System (INIS)

Pinard Legry, G.

1994-01-01

Titanium and zirconium pure and base alloys are protected by an oxide film with anionic vacancies which gives a very good resistance to corrosion in oxidizing medium, in some ph ranges. Results of pitting and crevice corrosion are given for Cl - , Br - , I - ions concentration with temperature and ph dependence, also with oxygenated ions effect. (A.B.). 32 refs., 6 figs., 3 tabs

Effect of Bi on the corrosion resistance of zirconium alloys

International Nuclear Information System (INIS)

Yao Meiyi; Zhou Bangxin; Li Qiang; Zhang Weipeng; Zhu Li; Zou Linghong; Zhang Jinlong; Peng Jianchao

2014-01-01

In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-l.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-l.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4 + xBi, S5 + xBi, T5 + xBi and Zr-1Nb + xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360 ℃/18.6 MPa and in superheated steam at 400 ℃/10.3 MPa. The microstructure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Microstructure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance. (authors)

Experimental and thermodynamic study of the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems

International Nuclear Information System (INIS)

Jourdan, J.

2009-11-01

This work is a contribution to the development of innovative concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of phase equilibria in the systems is essential prior to the implementation of such a promising solution. This study consisted in an experimental determination of the erbium-zirconium phase diagram. For this, we used many different techniques in order to obtain diagram data such as solubility limits, solidus, liquidus or invariant temperatures. These data allowed us to present a new diagram, very different from the previous one available in the literature. We also assessed the diagram using the CALPHAD approach. In the gadolinium-zirconium system, we determined experimentally the solubility limits. Those limits had never been determined before, and the values we obtained showed a very good agreement with the experimental and assessed versions of the diagram. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems. The first system has been investigated experimentally. The alloys fabrication has been performed using powder metallurgy. In order to obtain pure raw materials, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. The characterisation of the ternary pellets allowed the determination of two ternary isothermal sections at 800 and 1100 C. For the gadolinium-oxygen-zirconium system, we calculated the phase equilibria at temperatures ranging from 800 to 1100 C, using a homemade database compiled from literature assessments of the oxygen-zirconium, gadolinium-zirconium and gadolinia-zirconia systems. Finally, we determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of

Microstructural characterization of mechanically alloyed Al–Cu–Mn alloy with zirconium

Energy Technology Data Exchange (ETDEWEB)

Prosviryakov, A.S., E-mail: pro.alex@mail.ru; Shcherbachev, K.D.; Tabachkova, N.Yu.

2015-01-19

An evolution of Al–Cu–Mn alloy microstructure during its mechanical alloying with zirconium 20 wt% and after subsequent annealing was studied by X-ray diffraction, light microscopy and transmission electron microscopy. The effect of milling time on powder microhardness, Al lattice parameter, lattice microstrain and crystallite size was determined.

Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

International Nuclear Information System (INIS)

Grossman, L.N.; Packard, D.R.

1975-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

International Nuclear Information System (INIS)

Fedotov, A.

2003-01-01

The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

A study into the impact of interface roughness development on mechanical degradation of oxides formed on zirconium alloys

International Nuclear Information System (INIS)

Platt, P.; Wedge, S.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2015-01-01

As a cladding material used to encapsulate nuclear fuel pellets, zirconium alloys are the primary barrier separating the fuel and a pressurised steam or lithiated water environment. Degradation mechanisms such as oxidation can be the limiting factor in the life-time of the fuel assembly. Key to controlling oxidation, and therefore allowing increased burn-up of fuel, is the development of a mechanistic understanding of the corrosion process. In an autoclave, the oxidation kinetics for zirconium alloys are typically cyclical, with periods of accelerated kinetics being observed in steps of ∼2 μm oxide growth. These periods of accelerated oxidation are immediately preceded by the development of a layer of lateral cracks near the metal-oxide interface, which may be associated with the development of interface roughness. The present work uses scanning electron microscopy to carry out a statistical analysis of changes in the metal-oxide interface roughness between three different alloys at different stages of autoclave oxidation. The first two alloys are Zircaloy-4 and ZIRLO ™ for which analysis is carried out at stages before, during and after first transition. The third alloy is an experimental low tin alloy, which under the same oxidation conditions and during the same time period does not appear to go through transition. Assessment of the metal-oxide interface roughness is primarily carried out based on the root mean square of the interface slope known as the R dq parameter. Results show clear trends with relation to transition points in the corrosion kinetics. Discussion is given to how this relates to the existing mechanistic understanding of the corrosion process, and the components required for possible future modelling approaches

Phase Transformations in a Uranium-Zirconium Alloy containing 2 weight per cent Zirconium

Energy Technology Data Exchange (ETDEWEB)

Lagerberg, G

1961-04-15

The phase transformations in a uranium-zirconium alloy containing 2 weight percent zirconium have been examined metallographically after heat treatments involving isothermal transformation of y and cooling from the -y-range at different rates. Transformations on heating and cooling have also been studied in uranium-zirconium alloys with 0.5, 2 and 5 weight per cent zirconium by means of differential thermal analysis. The results are compatible with the phase diagram given by Howlett and Knapton. On quenching from the {gamma}-range the {gamma} phase transforms martensitically to supersaturated a the M{sub S} temperature being about 490 C. During isothermal transformation of {gamma} in the temperature range 735 to 700 C {beta}-phase is precipitated as Widmanstaetten plates and the equilibrium structure consists of {beta} and {gamma}{sub 1}. Below 700 C {gamma} transforms completely to Widmanstaetten plates which consist of {beta} above 660 C and of a at lower temperatures. Secondary phases, {gamma}{sub 2} above 610 C and {delta} below this temperature, are precipitated from the initially supersaturated Widmanstaetten plates during the isothermal treatments. At and slightly below 700 C the cooperative growth of |3 and {gamma}{sub 2} is observed. The results of isothermal transformation are summarized in a TTTdiagram.

Recrystallization resistance in aluminum alloys containing zirconium

International Nuclear Information System (INIS)

Ranganathan, K.

1991-01-01

Zirconium forms a fine dispersion of the metastable β' (Al 3 Zr) phase that controls recrystallization by retarding the motion of high-angle boundaries. The primary material chosen for this research was aluminum alloy 7150 containing zinc, magnesium, and copper as the major solute elements and zirconium as the dispersoid-forming element. The size, distribution, and the volume fraction of β' was controlled by varying the alloy composition and preheat practices. Preheated ingots were subjected to a specific sequence of hot-rolling operations to evaluate the resistance to recrystallization of the different microstructures. Optical and transmission electron microscopy (TEM) techniques were used to investigate the influence of dispersoid morphology resulting from the thermal treatments and deformation processing on the recrystallization behavior of the alloy. Studies were conducted to determine the influence of the individual solute elements present in 7150 on the precipitation of β' and consequently on the recrystallization behavior of the material. These studies were done on compositional variants of commercial 7150

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

Anisotropy of mechanical properties of zirconium and zirconium alloys

International Nuclear Information System (INIS)

Medrano, R.E.

1975-01-01

In studies of technological applications of zirconium to fuel elements of nuclear reactor, it was found that the use of plasticity equations for isotropic materials is not in agreement with experimental results, because of the strong anisotropy of zirconium. The present review describes recent progress on the knowledge of the influence of anisotropy on mechanical properties, after Douglass' review in 1971. The review was written to be selfconsistent, changing drastically the presentation of some of the referenced papers. It is also suggested some particular experiments to improve developments in this area

Corrosion resistant zirconium alloys prepared by powder metallurgy

International Nuclear Information System (INIS)

Wojeik, C.C.

1984-01-01

Pure zirconium and zirconium 2.5% niobium were prepared by powder metallurgy. The powders were prepared directly from sponge and consolidated by cold isostatic pressing and sintering. Hot isostatic pressing was also used to obtain full density after sintering. For pure zirconium the effects of particle size, compaction pressure, sintering temperature and purity were investigated. Fully densified zirconium and Zr-2.5%Nb exhibited tensile properties comparable to cast material at room temperature and 300 0 F (149 0 C). Pressed and sintered material having density of 94-99% had slightly lower tensile properties. Corrosion tests were performed in boiling 65% H/sub 2/SO/sub 4/, 70% HNO/sub 3/, 20% HCl and 20% HCl + 500 ppm FeCl/sub 3/ (a known pitting solution). For fully dense material the observed corrosion behavior was nearly equivalent to cast material. A slightly higher rate of attack was observed for samples which were only 94-99% dense. Welding tests were also performed on zirconium and Zr-2.5%Nb alloy. Unlike P/M titanium alloys, these materials had good weldability due to the lower content of volatile impurities in the powder. A slight amount of weld porosity was observed but joint efficiencies were always not 100%, even for 94-99% density samples. Several practical applications of the P/M processed material will be briefly described

Characteristics of Pilger Die Materials for Nuclear Zirconium Alloy Tubes

International Nuclear Information System (INIS)

Park, Ki Bum; Kim, In Kyu; Park, Min Young; Kahng, Jong Yeol; Kim, Sun Doo

2011-01-01

KEPCO Nuclear Fuel Company's (KEPCO NF) tube manufacturing facility, Techno Special Alloy (TSA) Plant, has started cold pilgering operation since 2008. It is obvious that the cold pilgering process is one of the key processes controlling the quality and the characteristics of the tubes manufactured, i.e. nuclear zirconium alloy tube in KEPCO NF. Cold pilgering is a rolling process for forming metal tubes in which diameter and wall thickness are reduced in a number of forming steps, using ring dies at outside of the tube and a curved mandrel at inside to reduce tube cross sections by up to 90 percent. The OD size of tube is reduced by a pair of dies, and ID size and wall thickness is controlled simultaneously by mandrel. During the cold pilgering process, both tools are the critical components for providing qualified tube. Development of pilger die and mandrel has been a significant importance in the zirconium tube manufacturing and a major goal of KEPCO NF. The objective of this study is to evaluate the life time of pilger die during pilgering. Therefore, a comparison of the heat treatment and mechanical properties of between AISI 52100 and AISI H13 materials was made in this study

The Development of an In-Situ TEM Technique for Studying Corrosion Behavior as Applied to Zirconium-Based Alloys

Science.gov (United States)

Harlow, Wayne

Zirconium-based alloys are a commonly used material for nuclear fuel rod cladding, due to its low neutron cross section and good corrosion properties. However, corrosion is still a limiting factor in fuel rod lifespan, which restricts burn up levels, and thus efficiency, that can be achieved. While long-term corrosion behavior has been studied through both reactor and autoclave samples, the oxide nucleation and growth behavior has not been extensively studied. This work develops a new technique to study the initial stages of corrosion in zirconium-based alloys and the microstructural effects on this process by developing an environmental cell system for the TEM. Nanoscale oxidation parameters are developed, as is a new FIB technique to support this method. Precession diffraction is used in conjunction with in-situ TEM to observe the initial stages of corrosion in these alloys, and oxide thickness is estimated using low-loss EELS. In addition, the stress stabilization of tetragonal ZrO 2 is explored in the context of sample preparation for TEM. It was found that in-situ environmental TEM using an environmental cell replicates the oxidation behavior observed in autoclaved samples in both oxide structure and phases. Utilizing this technique, it was shown that cracking of the oxide layer in zirconium-based alloys is related to oxide relaxation, and not thermal changes. The effect of secondary phase particles on oxidation behavior did not present significant results, however a new method for studying initial oxidation rates using low-loss EELS was developed.

Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

OpenAIRE

Bo Cheng; Young-Jin Kim; Peter Chou

2016-01-01

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

In situ Raman Spectroscopy of Oxide Films on Zirconium Alloy in Simulated PWR Primary Water Condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-05-15

The two layered oxide structure is formed in pre-transition oxide for the zirconium alloy in high temperature water environment. It is known that the corrosion rate is related to the volume fraction of zirconium oxide and the pores in the oxides; therefore, the aim of this paper is to investigate the oxidation behavior in the pretransition zirconium oxide in high-temperature water chemistry. In this work, Raman spectroscopy was used for in situ investigations for characterizing the phase of zirconium oxide. In situ Raman spectroscopy is a well-suited technique for investigating in detail the characteristics of oxide films in a high-temperature corrosion environment. In previous studies, an in situ Raman system was developed for investigating the oxides on nickel-based alloys and low alloy steels in high-temperature water environment. Also, the early stage oxidation behavior of zirconium alloy with different dissolved hydrogen concentration environments in high temperature water was treated in the authors' previous study. In this study, a specific zirconium alloy was oxidized and investigated with in situ Raman spectroscopy for 100 d oxidation, which is close to the first transition time of the zirconium alloy oxidation. The ex situ investigation methods such as transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDS) were used to further characterize the zirconium oxide structure. As oxidation time increased, the Raman peaks of tetragonal zirconium oxide were merged or became weaker. However, the monoclinic zirconium oxide peaks became distinct. The tetragonal zirconium oxide was just found near the O/M interface and this could explain the Raman spectra difference between the 30 d result and others.

Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

International Nuclear Information System (INIS)

Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

2007-01-01

Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

Impact of Zr + 2.5% Nb alloy corrosion upon operability of RBMK-1000 fuel channels

International Nuclear Information System (INIS)

Kovyrshin, V.; Zaritsky, N.

1999-01-01

The basic components of RBMK-1000 core (fuel channels, bimetal adapters, claddings of fuel elements, etc.) are of zirconium alloys. Their corrosion is one of factors influencing upon fuel channels operability. Dynamics of channel tubes nodular corrosion development is presented by the results of in-reactor investigation at ChNPP. Radiation-induced mechanism of corrosion damage of tubes surface in contact with coolant was formulated and substantiated by data of post-reactor studies. Within the certain time period of operation corrosion of zirconium alloy of lower bimetal adapter along with removal from there of corrosion products are predominant within the whole process of reactor elements corrosion. The experimental and calculating method was proposed and substantiated to predict time duration up to loss of fuel channels leak tightness. The approaches were generalized to control state of fuel channels material to assess their operability under operation of RBMK-1000 reactors. (author)

The degradation of zirconium alloys in nuclear reactors - a review

International Nuclear Information System (INIS)

Lim, D.; Graham, N.A.

1986-01-01

This report presents the findings of a survey of available non-Canadian literature on the oxidation and hydriding of zirconium alloys. Much of the literature was found to address the Zircaloys, particularly when used as fuel cladding subjected to a radioactive and oxidizing environment. Hydriding of Zircaloys is mainly attributed to oxidation. The survey revealed that Zr-Nb alloys have been included in some investigations; however, data on the long-term degradation of Zr-2.5 wt% Nb, in particular, were scarce. The reviewed literature did not lead to conclusions regarding the potential for accelerated hydriding due to corrosion at crevices and/or second-phase particles, nor did it lead to conclusions as to the potential for a 'breakaway' in oxidation and hydrogen acquisition in long service life of Zr-Nb alloys. Specific information on service experience in U.S.S.R. power reactors could not be obtained; however, most of the information surveyed leads to the conclusion that fuel channels having Zr-2.5 wt% Nb pressure tubes should perform satisfactorily with respect to degradation from corrosion and hydriding provided they are installed correctly and are not operated under conditions that are far removed from those anticipated in design. 91 refs

Hydrogen charging, hydrogen content analysis and metallographic examination of hydride in zirconium alloys

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Mukherjee, S.; Roychowdhury, S.; Srivastava, D.; Sinha, T.K.; De, P.K.; Banerjee, S.; Gopalan, B.; Kameswaran, R.; Sheelvantra, Smita S.

2003-12-01

Gaseous and electrolytic hydrogen charging techniques for introducing controlled amount of hydrogen in zirconium alloy is described. Zr-1wt%Nb fuel tube, zircaloy-2 pressure tube and Zr-2.5Nb pressure tube samples were charged with up to 1000 ppm of hydrogen by weight using one of the aforementioned methods. These hydrogen charged Zr-alloy samples were analyzed for estimating the total hydrogen content using inert gas fusion technique. Influence of sample surface preparation on the estimated hydrogen content is also discussed. In zirconium alloys, hydrogen in excess of the terminal solid solubility precipitates out as brittle hydride phase, which acquire platelet shaped morphology due to its accommodation in the matrix and can make the host matrix brittle. The F N number, which represents susceptibility of Zr-alloy tubes to hydride embrittlement was measured from the metallographs. The volume fraction of the hydride phase, platelet size, distribution, interplatelet spacing and orientation were examined metallographically using samples sliced along the radial-axial and radial-circumferential plane of the tubes. It was observed that hydride platelet length increases with increase in hydrogen content. Considering the metallographs generated by Materials Science Division as standard, metallographs prepared by the IAEA round robin participants for different hydrogen concentration was compared. It is felt that hydride micrographs can be used to estimate not only that approximate hydrogen concentration of the sample but also its size, distribution and orientation which significantly affect the susceptibility to hydride embrittlement of these alloys. (author)

Towards an understanding of zirconium alloy corrosion

International Nuclear Information System (INIS)

Cox, B.

1976-08-01

A brief historical summary is given of the development of a programme for understanding the corrosion mechanisms operating for zirconium alloys. A general summary is given of the progress made, so far, in carrying through this programme. (author)

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Energy Technology Data Exchange (ETDEWEB)

Michael V. Glazoff

2013-08-01

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

Control of microstructure during hot working of zirconium alloys

International Nuclear Information System (INIS)

Chakravartty, J.K.; Banerjee, S.

2005-01-01

Hot working is considered to be the most important step involved in the fabrication of zirconium alloys for nuclear reactor applications for two reasons: i) the scale of the microstructure and texture of the final product is decided at this stage and ii) the hot deformed microstructure provides a suitable starting microstructure for the subsequent fabrication steps. The resultant microstructure in turn controls the properties of the final product. In order to obtain final product with a suitable microstructure and with specified mechanical properties on a repeatable basis the control of microstructure during hot working is of paramount importance. This is usually done by studying the constitutive behaviour of the material under hot working conditions and by constructing processing maps. In the latter method, strain rate sensitivity is mapped as a function of temperature and strain rate to delineate domains within the bounds of which a specific deformation mechanism dominates. Detail microstructural analysis is then carried out on the samples deformed within the domains. Using this methodology, processing maps have been constructed for various zirconium alloys. These maps have been found to be very useful for optimizing the hot workability and control of microstructure of zirconium alloys. (author)

Thermo-mechanical treatment of zirconium alloys

International Nuclear Information System (INIS)

Levy, I.S.

1975-01-01

A zirconium alloy comprising at least 95 percent Zr (Zircaloy), which has been thoroughly annealed, is greatly increased in strength without substantial loss in ductility by subjecting it to tensile creep deformation in a temperature range in which creep will occur, yet which is below the temperature for significant recovery. (U.S.)

Nuclear fuel element and container

International Nuclear Information System (INIS)

Grubb, W.T.; King, L.H.

1981-01-01

The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

Thermal creep behavior of N36 zirconium alloy cladding tube

International Nuclear Information System (INIS)

Wang, P.; Zhao, W.; Dai, X.

2015-01-01

N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

The technologies of zirconium production for nuclear fuel components in Ukraine

International Nuclear Information System (INIS)

Semenov, G.R.

2000-01-01

Perspectives of development zirconium alloys and WWER-1000 assemble components production in Ukraine are considered. Basic technological production processes of zirconium alloys in conditions of Ukrainian enterprises and modern requirements are analyzed. The critical processes on technical and economic criteria are defined. The main directions of activity and steps on technological processes improvement for production quality providing are offered. (author)

Characterization of zirconium alloy oxidation films by alternating current impedance

International Nuclear Information System (INIS)

Rosecrans, P.M.

1984-01-01

Kinetics of zirconium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

High corrosion-resistant fuel spacers

International Nuclear Information System (INIS)

Yoshida, Toshimi; Takase, Iwao; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

1986-01-01

Purpose: To enable manufacturing BWR fuel spacers by prior-art production process, using a zirconium-base alloy having very excellent corrosion resistance. Method: A highly improved nodular-resistant, corrosion-resistant zirconium alloy is devised by adding a slight amount of niobium, titanium and vanadium to zircaloy, of which fuel spacers are produced. That is, there can be obtained an alloy having much more excellent nodular resistance than conventional zircaloy, and free from a large change in plasticity, workability, and weldability, by adding to zirconium about 1.5 % of tin, about 0.15 % of iron, about 0.05 % of chromium, about 0.05 % of nickel, and 0.05 to 0.5 % of at least one or two kinds of niobium, titanium and vanadium. Using this zirconium-base alloy can manufacture fuel spacers by the same manufacturing process, thus improving economy and reliability. (Kamimura, M.)

FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

International Nuclear Information System (INIS)

Lott, Randy G.

2003-01-01

OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

International Nuclear Information System (INIS)

Rebak, Raul B.

2014-01-01

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

2014-09-30

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

International Nuclear Information System (INIS)

McDeavitt, Sean M.

2011-01-01

beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A - MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled 'Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled 'Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors' Appendix B - External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, 'Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, 'Uranium Powder Production Using a Hydride-Dehydride Process,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C - Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys' presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow

Round robin test for zirconium alloys in 400 deg C steam: results from EDF

International Nuclear Information System (INIS)

Blat, M.

1994-01-01

The EDF Material Studies Branch has participated in the Round Robin program of uniform corrosion on zirconium alloys. The objectives of these Round Robin corrosion tests are to generate new uniform corrosion weight gain date utilizing modern zirconium alloy products and to improve the International and ASTM standards. (author). 2 tabs., 7 appendix., 2 refs

Collaborative analysis for certification of zirconium and zirconium base alloy reference materials JAERI-Z11 to Z16

International Nuclear Information System (INIS)

1985-03-01

The second Sub-Committee on Zircaloy Analysis was organized in April 1978, under the Committee on Analytical Chemistry on Nuclear Fuels and Reactor Materials, JAERI, for the renewal of zirconium and zirconium base alloy certified reference materials (CRMs). The Sub-Committee carried out collaborative analysis among 13 participating laboratories for the certification of the CRMs, JAERI-Z11 to Z18, after development, improvement and evaluation of analytical methods during the period of May 1978 to June 1982. As the result of the collaborative analysis, the certified value was given for 18 elements (Sn, Fe, Ni, Cr, B, Cd, U, Cu, Co, Mn, Pb, Al, Ti, Si, Mo, W, Hf, C) in the CRMs. The first part of this report includes general discussion, the second part principles of certification, the third part development and verification of analytical methods, and the fourth part evaluation of analytical results on 17 elements. Preparation of Z11 to Z18, and certification for carbon in JAERI-Z17 and Z18 were reported separately in JAERI-M 83-241 and M 83-035, respectively. (author)

Influence of impurities and ion surface alloying on the corrosion resistance of E110 alloy

International Nuclear Information System (INIS)

Kalin, B. A.; Volkov, N. V.; Valikov, R. A.; Novikov, V. V.; Markelov, V. A.; Pimenov, Yu. V.

2013-01-01

The corrosion resistance of zirconium alloys depends on their structural-phase state, the type of core coolant and operating factors. The formation of a protective oxide film on the zirconium alloys is sensitive to the content of impurity atoms present in the charge base of alloys and accumulating in them in the manufacture of products. The impurity composition of the initial zirconium is determined by the method of its manufacture and generally remains unchanged in the products, deter-mining their properties, including their corrosion resistance. An increased content of impurities (C, N, Al, Mo, Fe) both individually and in their combination negatively affects the corrosion resistance of zirconium and its alloys. One of the potentially effective methods to increase the protective properties of oxide films on zirconium alloys is a surface alloying using the regime of mixing the atoms of a film, preliminarily coated on the surface, and the atoms of a target. This method makes it possible to form a given structural-phase state in the thin surface layer with unique physicochemical properties and thus to in-crease the corrosion resistance and wear resistance of fuel claddings. In this context, the object of investigation was samples of cladding tubes from alloy E110 with various content of impurity elements (nitrogen, aluminum, and carbon) with the aim to reduce the negative influence of impurities on the corrosion resistance by changing the structural-phase state of the surface layer of fuel claddings and fuel assembly components with alloying in the regime of ion mixing of atoms

Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

International Nuclear Information System (INIS)

Turque, Isabelle

2016-01-01

Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

Prospects for zirconium structural alloys at high temperatures

International Nuclear Information System (INIS)

Thomas, W.R.

1969-05-01

Improved station efficiencies and lower capital costs provide incentives for the development of zirconium alloys for pressure tubes which can operate at temperatures above 450 o C. The experience of the Ti industry indicates that a complex alloy containing solution hardeners of Sn or Al and precipitation hardeners of Mo and Nb and perhaps Si will be required. The thermal neutron cross-section of the alloy will be about 10% higher than Zircaloy-2 and because of its poor corrosion resistance will require cladding with a corrosion resistant alloy such as Zr-Cr. Results to date indicate that such a pressure tube is feasible. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1975-01-01

The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1974-01-01

The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

McDeavitt, Sean M

2011-04-29

outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

Fretting wear behavior of zirconium alloy in B-Li water at 300 °C

Science.gov (United States)

Zhang, Lefu; Lai, Ping; Liu, Qingdong; Zeng, Qifeng; Lu, Junqiang; Guo, Xianglong

2018-02-01

The tangential fretting wear of three kinds of zirconium alloys tube mated with 304 stainless steel (SS) plate was investigated. The tests were conducted in an autoclave containing 300 °C pressurized B-Li water for tube-on-plate contact configuration. The worn surfaces were examined with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and 3D microscopy. The cross-section of wear scar was examined with transmission electron microscope (TEM). The results indicated that the dominant wear mechanism of zirconium alloys in this test condition was delamination and oxidation. The oxide layer on the fretted area consists of outer oxide layer composed of iron oxide and zirconium oxide and inner oxide layer composed of zirconium oxide.

Anticorrosion ion implantation of fragments of zirconium fuel can specimens

International Nuclear Information System (INIS)

Kalin, B.A.; Osipov, V.V.; Volkov, N.V.; Khernov, V.Yu.

2001-01-01

Aimed at the study of specific features of oxide film formation in the initial stage of Eh110 and Eh635 alloy fuel can oxidation the modification of tubular specimen surfaces is performed using an ion mixing technique, and the structure of oxide films produced in a steam-water environment is investigated. Using the method of vacuum vapor deposition the outer surface of specimens is coated with alloying element films irradiated by a polyenergetic Ar + ion beam with a 10 keV mean energy up to radiation doses of (7-10) x 10 17 ion/cm 2 . Monatomic (Al, Fe, Cu, Cr, Mo, Sn) or diatomic (Al-Fe, Al-Mo, Al-Sn, Fe-Cu, Fe-Mo, Fe-Sn, Cr-Mo, Cr-Sn) implantation into a zirconium cladding occurs under irradiation effect. The positive influence of combined intrusion of Al and other elements is revealed. The presence of Al atoms enhances the oxide film structure. The least ZeO 2 film thickness is observed when alloying with molybdenum, Al-Fe, Al-Mo and Al-Sn [ru

High energy beam thermal processing of alpha zirconium alloys and the resulting articles

International Nuclear Information System (INIS)

Sabol, G.P.; McDonald, S.G.; Nurminen, J.I.

1983-01-01

Alpha zirconium alloy fabrication methods and resultant products exhibiting improved high temperature, high pressure steam corrosion resistance. The process, according to one aspect of this invention, utilizes a high energy beam thermal treatment to provide a layer of beta treated microstructure on an alpha zirconium alloy intermediate product. The treated product is then alpha worked to final size. According to another aspect of the invention, high energy beam thermal treatment is used to produce an alpha annealed microstructure in a Zircaloy alloy intermediate size or final size component. The resultant products are suitable for use in pressurized water and boiling water reactors

Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

Directory of Open Access Journals (Sweden)

Bo Cheng

2016-02-01

Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

The Hydrogen Pickup Behavior for Zirconium-based Alloys in Various Out-of-pile Corrosion Test Conditions

Energy Technology Data Exchange (ETDEWEB)

Aomi, M.; Etoh, Y.; Ishimoto, S.; Une, K. [Nippon Nuclear Fuel Development, Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki-ken, 311-1313 (Japan); Ito, K. [Global Nuclear Fuel Japan Co., Ltd., 3-1 Uchikawa 2-chome, Yokosuka-shi, Kanagawa-ken, 239-0836 (Japan)

2009-06-15

An acceleration of hydrogen absorption in zirconium alloy claddings at high burnups is one of the most important issues limiting the fuel performance from the viewpoint of cladding integrity. In this context, advanced cladding materials with higher corrosion resistant and lower hydrogen absorption properties have been widely searched in various organizations. In this study, four kinds of zirconium-based alloys, whose in-pile data had been acquired [1,2] were subjected to comprehensive out-of-pile corrosion tests with various temperature and atmosphere conditions in order to investigate the correlation between in-pile and out-of-pile corrosion and hydrogen pick-up behavior, i.e. Zry-2, GNF-Ziron (Zry-2-based alloy with {approx}0.25 wt % of Fe), Hi-FeNi Zircaloy (Zry-2-based alloy with {approx}0.25 wt % of Fe and {approx}0.1 wt% Ni), and VB (Zr-based alloy containing Sn, Cr, and {approx}0.5 wt % of Fe). All the alloys were annealed in RXA condition. The out-of-pile corrosion tests were carried out in three different conditions of 400 deg. C steam, 475 deg. C supercritical water, and 290 deg. C LiOH aqueous solution. In addition to these alloys, several Zry-2-based alloys with various iron contents were tested in 290 deg. C LiOH aqueous solution. Among the four corrosion conditions, the 290 deg. C LiOH aqueous solution test well screened the hydrogen pick-up behavior of the alloys. The hydrogen absorption decreased with higher iron contents in the alloys in both the out-of-pile and in-pile conditions. Especially, the distinct suppression of hydrogen absorption was observed for VB with the highest iron content. The similar dependence of iron content on the hydrogen pick-up fraction was also obtained for the Zry-2-based alloys with different iron contents, which were corroded in the 290 deg. C LiOH aqueous solution condition. As for the corrosion behavior in the 290 deg. C LiOH aqueous solution condition, the weight gains of Zry-2, GNF-Ziron and VB followed the 1

Corrosion of zirconium alloys in alternating pH environment

International Nuclear Information System (INIS)

Mayer, P.; Manolescu, A.V.

1985-01-01

Behaviour of two commercial alloys, Zircaloy-2 and zirconium-2.5 wt% niobium were investigated in an environment of alternating pH. Corrosion advancement and scale morphology of coupons exposed to aqueous solution of LiOH (pH 10.2 and 14) were followed as a function of temperature (300-360 degreesC) and time (up to 165 days). The test sequence consisted of short term exposure to high pH and re-exposure to low pH solutions for extended period of time followed by a short term test in high pH. The results of these tests and detailed post-corrosion analysis indicate a fundamental difference between the corrosion behaviour of these two materials. Both alloys corrode fast in high pH environments, but only zirconium-2.5 wt% niobium continues to form detectable new oxide in low pH solution

Proceedings [of the] symposium on zirconium alloys for reactor components

International Nuclear Information System (INIS)

1992-01-01

A two day symposium on zirconium alloys for reactor components (ZARC-91) was organised during 12-13, 1991. There were 6 invited talks and 43 contributed papers in 10 technical sessions. This symposium, took stock of the progress achieved in the development, design, fabrication and quality assurance of zirconium alloy components and emphasized the R and D efforts required for meeting the challenges posed by the rapid growth of nuclear power in our country. Topics like physical metallurgy, corrosion and irradiation behaviour, and in-service inspection were also covered. The proceedings/papers are arranged under the headings: (1)invited talks, (2)fabrication, (3)design requirement, (4)quality assurance, (5)irradiation damage and PIE, (6)corrosion and hydriding, and (7)in-service inspection. (N.B.). refs., figs., tabs

Technology readiness level (TRL) assessment of cladding alloys for advanced nuclear fuels

International Nuclear Information System (INIS)

Shepherd, Daniel

2015-01-01

Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic (F/M), reduced activation (RA) and oxide dispersion strengthened (ODS) steels as well as advanced iron-based alloys (Kanthal alloys). Also assessed are alloys of zirconium, nickel (including Hastelloy R ), titanium, chromium, vanadium and refractory metals (Nb, Mo, Ta and W). Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. (authors)

Accelerated irradiation growth of zirconium alloys

International Nuclear Information System (INIS)

Griffiths, M.; Gilbert, R.W.; Fidleris, V.

1989-01-01

This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

Autoclave Testing on Zirconium Alloy Materials

International Nuclear Information System (INIS)

Hoffmann, Petra-Britt; Sell, Hans-Juergen; Garzarolli, Friedrich

2012-09-01

The corrosion of Zirconium components like fuel rod claddings and spacer grids is limiting lifetime and duty of these components. In Pressurized and Boiling Water Reactors (PWR and BWR), different corrosion phenomena are of interest. Although in-pile experience is the final proof for a material development, significant experience was gained by autoclave tests, trying to simulate in-pile conditions but reducing time for return of experience by increased temperatures. For PWR application, the uniform corrosion is studied in water at up to 370 deg. C and in high pressure steam at 400 deg. C, and for BWR, the nodular corrosion is studied in high pressure steam at 500-520 deg. C. Particular attention has to be given to the corrosion media, because oxidative traces in the water can significantly affect the corrosion response. An extensive air removal is thus important for all corrosion tests. This links to the different water chemistry conditions that have been investigated as separate effects otherwise difficult to separate under in-pile conditions. Uniform corrosion in 350 deg. C water is usually a cyclic process with repeated rate transitions. In addition, at high exposure times an acceleration of corrosion can occur, e.g. for Zr-Sn alloys with a high Sn content. In 400 deg. C steam, corrosion rate decreases somewhat with increasing time. Uniform corrosion rate of Zr alloys depends on their Sn- and Fe+Cr contents as well as on their annealing parameters with a similar trend as in PWR and on their yield strength, however with an opposite trend compared to BWR conditions. Nodular corrosion of BWR alloys depends on the annealing parameter with a similar trend as in PWR and out-of-reactor also significantly on the Fe+Cr content. The hydrogen pickup fraction (HPUF) depends largely on details of the water chemistry and can particularly depend on autoclave degassing and probably also on autoclave contaminations. Thus any HPUF value from out-of- pile corrosion tests is only

Fuel element

International Nuclear Information System (INIS)

Armijo, J.S.

1976-01-01

A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

Oxidation of zirconium alloys in steam: influence of tetragonal zirconia on oxide growth mechanism

International Nuclear Information System (INIS)

Godlewski, J.

1990-07-01

The oxidation of zirconium alloys in presence of steam, presents after a 'parabolic' growth law, an acceleration of the oxidation velocity. This phenomenon limits the use of zirconium alloys as nuclear fuel cladding element. In order to determine the physico-chemical process leading to this kinetic transition, two approaches have been carried out: the first one has consisted to determine the composition of the oxide layer and its evolution with the oxidation time; and the second one to determine the oxygen diffusion coefficients in the oxide layers of pre- and post-transition as well as their evolution with the oxidation time. The composition of the oxide layers has been determined by two analyses techniques: the X-ray diffraction and the laser Raman spectroscopy. This last method has allowed to confirm the presence of tetragonal zirconium oxide in the oxide layers. Analyses carried out by laser Raman spectroscopy on oxides oblique cuttings have revealed that the tetragonal zirconium oxide is transformed in monoclinic phase during the kinetic transition. A quantitative approach has allowed to corroborate the results obtained by these two techniques. In order to determine the oxygen diffusion coefficients in the oxides layers, two diffusion treatments have been carried out: 1)under low pressure with D 2 18 O 2 ) under high pressure in an autoclave with H 2 18 O. The oxygen 18 concentration profiles have been obtained by two analyses techniques: the nuclear microprobe and the secondary ions emission spectroscopy. The obtained profiles show that the mass transport is made by the volume and particularly by the grain boundaries. The corresponding diffusion coefficients have been calculated with the WHIPPLE and LE CLAIRE solution. The presence of tetragonal zirconium oxide, its relation with the kinetic transition, and the evolution of the diffusion coefficients with the oxidation time, are discussed in terms of internal stresses in the oxide layer and of the oxide layer

Method for forming nuclear fuel containers of a composite construction and the product thereof

International Nuclear Information System (INIS)

Cheng, B.-C.; Rosenbaum, H.S.; Armijo, J.S.

1981-01-01

An improved method of producing a composite nuclear fuel container is described which comprises a casing or fuel sheath of zirconium or its alloy with a lining cladding of deposited copper superimposed over the inside surface of the zirconium or alloy and a layer of oxide of the zirconium or alloy formed on the inside surface of the casing or sheath. (U.K.)

Traps in Zirconium Alloys Oxide Layers

Directory of Open Access Journals (Sweden)

Helmar Frank

2005-01-01

Full Text Available Oxide films long-time grown on tubes of three types of zirconium alloys in water and in steam were investigated, by analysing I-V characteristic measured at constant voltages with various temperatures. Using theoretical concepts of Rose [3] and Gould [5], ZryNbSn(Fe proved to have an exponential distribution of trapping centers below the conduction band edge, wheras Zr1Nb and IMP Zry-4 proved to have single energy trap levels.

The oxidation kinetics of zirconium alloys applicable to loss-of-coolant accidents

International Nuclear Information System (INIS)

Parsons, P.D.; Miller, W.N.

1977-10-01

A review is presented of the available published measurements of the rate of reaction between zirconium alloys and steam and, in some cases, oxygen. Attempts are made to define from all the experimental data a suitable rate equation which is appropriate over the range of temperatures relevant to LOCA conditions. The data reviewed encompass a temperature range 910 0 C to the melting point of zirconium, 1852 0 C. It can be concluded that within 910 to 1577 0 C, Zircaloy-2, Zircaloy-4 and Zr/2 1/2%Nb alloys have the same response to oxidation. (author)

Evolution of zirconium-based precipitates during oxidation and irradiation of Zr alloys (impact on the oxidation kinetics of Zr alloys)

International Nuclear Information System (INIS)

Pecheur, Dominique

1993-01-01

As the oxidation of the zircaloy sheath is one of the factors which limit the lifetime of nuclear fuel rods, this research thesis aims at a better knowledge of the involved oxidation mechanisms and to improve the oxidation resistance in order to increase rod lifetime. Oxidation test performed in autoclave to study zirconium alloy oxidation without irradiation showed that oxidation kinetics is significantly higher under irradiation. This difference is attributed to a different evolution of the sheath material under irradiation. Thus, this research focused on the role of precipitates in the oxidation process of zirconium alloys, and on the impact of their amorphization on this oxidation. After a detailed description of the context and of the various implemented experimental means, the author presents the results obtained on a reference material on the one hand, and on a material irradiated by ions or neutrons on the other hand. More particularly, the author studied in these both cases the introduction of precipitates in the oxide layer by transmission electronic microscopy, and oxidation kinetics obtained in autoclave on these two types of material. He reports the analysis of the introduction of precipitates in the oxide layer formed on the reference material. He proposes interpretations for the evolutions of structure and of chemical compositions of precipitates in the oxide layer. These observations are then correlated with oxidation kinetics in these alloys. Finally, the author discusses results of oxidation tests obtained on materials irradiated by ions and by neutrons [fr

High-intensity low energy titanium ion implantation into zirconium alloy

Science.gov (United States)

Ryabchikov, A. I.; Kashkarov, E. B.; Pushilina, N. S.; Syrtanov, M. S.; Shevelev, A. E.; Korneva, O. S.; Sutygina, A. N.; Lider, A. M.

2018-05-01

This research describes the possibility of ultra-high dose deep titanium ion implantation for surface modification of zirconium alloy Zr-1Nb. The developed method based on repetitively pulsed high intensity low energy titanium ion implantation was used to modify the surface layer. The DC vacuum arc source was used to produce metal plasma. Plasma immersion titanium ions extraction and their ballistic focusing in equipotential space of biased electrode were used to produce high intensity titanium ion beam with the amplitude of 0.5 A at the ion current density 120 and 170 mA/cm2. The solar eclipse effect was used to prevent vacuum arc titanium macroparticles from appearing in the implantation area of Zr sample. Titanium low energy (mean ion energy E = 3 keV) ions were implanted into zirconium alloy with the dose in the range of (5.4-9.56) × 1020 ion/cm2. The effect of ion current density, implantation dose on the phase composition, microstructure and distribution of elements was studied by X-ray diffraction, scanning electron microscopy and glow-discharge optical emission spectroscopy, respectively. The results show the appearance of Zr-Ti intermetallic phases of different stoichiometry after Ti implantation. The intermetallic phases are transformed from both Zr0.7Ti0.3 and Zr0.5Ti0.5 to single Zr0.6Ti0.4 phase with the increase in the implantation dose. The changes in phase composition are attributed to Ti dissolution in zirconium lattice accompanied by the lattice distortions and appearance of macrostrains in intermetallic phases. The depth of Ti penetration into the bulk of Zr increases from 6 to 13 μm with the implantation dose. The hardness and wear resistance of the Ti-implanted zirconium alloy were increased by 1.5 and 1.4 times, respectively. The higher current density (170 mA/cm2) leads to the increase in the grain size and surface roughness negatively affecting the tribological properties of the alloy.

Solute redistribution studies in oxidised zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Khera, S K; Kale, G B; Gadiyar, H S [Bhabha Atomic Research Centre, Bombay (India). Metallurgy Div.

1977-01-01

Electron microprobe studies on solute distribution in oxide layers and in the regions near oxide metal interface have been carried out in the case of zircaloy-2 and zirconium binary alloys containing niobium, tin, iron, copper, chromium and nickel and oxidised in steam at 550 deg C. In the case of alloys having higher oxidation rates, the oxide of solute element was found to dissolve in ZrO/sub 2/ without any composition variation. However, for solute addition with limited solubility like Cr, Cu and Fe, solute enrichment at metal/oxide interface and depletion of the same matrix has been observed. The intensity profiles for nickel distribution were also found to be identical to Fe or Cr distribution. The mode of solute distribution has been discussed in relation to oxidation behaviour of these alloys.

Plate-shaped transformation products in zirconium-base alloys

International Nuclear Information System (INIS)

Banerjee, S.; Dey, G.K.; Srivastava, D.

1997-01-01

Plate-shaped products resulting from martensitic, diffusional, and mixed mode transformations in zirconium-base alloys are compared in the present study. These alloys are particularly suitable for the comparison in view of the fact that the lattice correspondence between the parent β (bcc) and the product α (hcp) or γ-hydride (fct) phases are remarkably similar for different types of transformations. Crystallographic features such as orientation relations, habit planes, and interface structures associated with these transformations have been compared, with a view toward examining whether the transformation mechanisms have characteristic imprints on these experimental observables

Electrochemical impedance spectroscopic study of passive zirconium

Energy Technology Data Exchange (ETDEWEB)

Ai Jiahe; Chen Yingzi [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Urquidi-Macdonald, Mirna [Department of Engineering Science and Mechanics, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D. [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States)], E-mail: ddm2@psu.edu

2008-09-30

Spent, unreproccessed nuclear fuel is generally contained within the operational fuel sheathing fabricated from a zirconium alloy (Zircaloy 2, Zircaloy 4, or Zirlo) and is then stored in a swimming pool and/or dry storage facilities until permanent disposal in a licensed repository. During this period, which begins with irradiation of the fuel in the reactor during operation, the fuel sheathing is exposed to various, aggressive environments. The objective of the present study was to characterize the nature of the passive film that forms on pure zirconium in contact with an aqueous phase [0.1 M B(OH){sub 3} + 0.001 M LiOH, pH 6.94] at elevated temperatures (in this case, 250 deg. C), prior to storage, using electrochemical impedance spectroscopy (EIS) with the data being interpreted in terms of the point defect model (PDM). The results show that the corrosion resistance of zirconium in high temperature, de-aerated aqueous solutions is dominated by the outer layer. The extracted model parameter values can be used in deterministic models for predicting the accumulation of general corrosion damage to zirconium under a wide range of conditions that might exist in some repositories.

Study of diffusion processes in the oxide layer of zirconium alloys

Directory of Open Access Journals (Sweden)

Sialini P.

2016-03-01

Full Text Available In the active zone of a nuclear reactor where zirconium alloys are used as a coating material, this material is subject to various harmful impacts. During water decomposition reactions, hydrogen and oxygen are evolved that may diffuse through the oxidic layer either through zirconium dioxide (ZrO2 crystals or along ZrO2 grains. The diffusion mechanism can be studied using the Ion Beam Analysis (IBA method where nuclear reaction 18O(p,α15N is used. A tube made of zirconium alloy E110 (with 1 wt. % of Nb was used for making samples that were pre-exposed in UJP PRAHA a.s. and subsequently exposed to isotopically cleansed environment of H2 18O medium in an autoclave. The samples were analysed with gravimetric methods and IBA methods performed at the electrostatic particle accelerator Tandetron 4130 MC in the Nucler Physics Institute of the CAS, Řež. With IBA methods, the overall thicknesses of corrosion layers on the samples, element composition of the alloy and distribution of oxygen isotope 18O in the corrosion layer and its penetration in the alloy were identified. The retrieved data shows at the oxygen diffusion along ZrO2 grains because there are two peaks of 18O isotope concentrations in the corrosion layer. These peaks occur at the environment-oxide and oxide-metal interface. The element analysis identified the presence of undesirable hafnium.

Investigation of in-pile grown corrosion films on zirconium-based alloys

International Nuclear Information System (INIS)

Gebhardt, O.; Hermann, A.; Bart, G.; Blank, H.; Ray, I.L.F.

1996-01-01

In-pile grown corrosion films on different fuel rod claddings (standard Zircaloy-4, extra low tin Zircaloy (ELS), and Zr2.5Nb) have been studied using a variety of experimental techniques. The aim of the investigations was to find out common features and differences between the corrosion layers grown on zirconium alloys having different composition. Methods applied were scanning and transmission electron microscopy (SEM, TEM), electrochemical impedance spectroscopy (EIS), and electrochemical anodization. The morphological differences have been observed between the specimens that could explain the irradiation enhancement of corrosion of Zircaloy-4. The features of the compact oxide close to the oxide/metal interface have been characterized by electrochemical methods. The relationship between the thickness of this protective oxide and the overall oxide thickness has been investigated by EIS. It was found that this relation is dependent on the location of the oxide along the fuel rod and on the corrosion rate

Amine extraction of lead(II) and zirconium(IV) with succinate media

International Nuclear Information System (INIS)

Mahamuni, S.V.; Mane, C.P.; Sargar, B.M.; Rajmane, M.M.; Anuse, M.A.

2004-01-01

Lead is an important constituent of various alloys, which are in increasing demand in manufacture of batteries and nuclear shielding while the use of zirconium in nuclear power plants as entirely cladding uranium fuel is most important. This study was carried out to optimize the extraction conditions for Pb(II) and zirconium(IV)

Methods for determination of zirconium in titanium alloys

International Nuclear Information System (INIS)

1985-01-01

Two methods for determining zirconium content in titanium alloys are specified in this standard. One is the ion-exchange/mandelic acid gravimetry for Zr content below 20 % down to 1 % while the other is the mandelic acid gravimetry for Zr content below 20 % down to 0.5 %. In the former, a specimen is decomposed by hydrochloric acid and hydrofluoric acid. After substances such as titanium are oxidized by adding nitric acid, the liquid is adjusted into a 4N hydrochloric acid - gN hydrofluoric acid solution, which is them passed through an ion-exchange column. The niobium and tantalum contents are absorbed while the titanium and zirconium contents flow out. Perchloric acid and sulfuric acid are poured in the solution to remove hydrofluoric acid. Aqueous ammonia is added to produce hydroxide of titanium and zirconium, which is then filtered out. The hydroxyde is dissolved in hydrochloric acid, and mandelic acid is poured to precipitate the zirconium content. The precipitate is ignited and the weight of the oxide formed is measured. The coprecipitated titanium content is determined by the absorptiometric method using hydrogen peroxide. Finally, the weight of the oxide is corrected. In the latter determination method, on the other hand, only several steps of the above procedure are used, namely, decomposition by hydrochloric acid, precipitation of zirconium, ignition of precipitate, measurement of oxide weight and weight correction. (Nogami, K.)

Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963)

International Nuclear Information System (INIS)

Baque, P.; Dominget, R.; Bossard, J.

1963-01-01

Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO 2 -cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [fr

Identification and characterization of a new Zirconium hydride

International Nuclear Information System (INIS)

Zhao, Z.

2007-01-01

In order to control the integrity of the fuel clad, alloy of zirconium, it is necessary to predict the behavior of zirconium hydrides in the environment (temperature, stress...), at a microscopic scale. A characterization study by TEM of hydrides has been realized. It shows little hydrides about 500 nm, in hydride Zircaloy 4. Then a more detailed study identified a new hydride phase presented in this paper. (A.L.B.)

Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

Energy Technology Data Exchange (ETDEWEB)

Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon [KAERI, Daejeon (Korea, Republic of)

2016-09-15

The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr{sub 2}O{sub 3}, and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

International Nuclear Information System (INIS)

Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon

2016-01-01

The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr 2 O 3 , and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged

Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

International Nuclear Information System (INIS)

Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R.

2017-01-01

The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples

Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

Energy Technology Data Exchange (ETDEWEB)

Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R. [Bhabha Atomic Research Centre, Mumbai (India)

2017-04-15

The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples.

PCI resistant light water reactor fuel cladding

International Nuclear Information System (INIS)

Foster, J.P.; Sabol, G.P.

1988-01-01

A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components Delayed Hydride Cracking

CERN Document Server

Puls, Manfred P

2012-01-01

By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the focus lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals.   This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing that our understanding of DHC is supported by progress across a broad range of fields. These include hysteresis associated with first-order phase transformations; phase relationships in coherent crystalline metallic...

Diffusion model of delayed hydride cracking in zirconium alloys

NARCIS (Netherlands)

Shmakov, AA; Kalin, BA; Matvienko, YG; Singh, RN; De, PK

2004-01-01

We develop a method for the evaluation of the rate of delayed hydride cracking in zirconium alloys. The model is based on the stationary solution of the phenomenological diffusion equation and the detailed analysis of the distribution of hydrostatic stresses in the plane of a sharp tensile crack.

Effects of ion implantation on corrosion of zirconium and zirconium base alloys

International Nuclear Information System (INIS)

Zelenskij, V.F.; Petel'guzov, I.A.; Rekova, L.P.; Rodak, A.G.

1989-01-01

The influence of He and Ar ion bombardment on the corrosion of Zr and Zr-1%Nb and Zr-2.5%Nb alloys is investigated with the aims of finding the irradiation influence laws, obtaining the dependences of the effect of increasing the corrosiuon resistance on the type and dose of bombarding ions and of finding the conditions for the maximum effect. The prolonged corrosion test of specimens (3500 hours) have shown that the strongest effect is obtained for the irradiation with Ar ions up to the dose 1x10 16 ion/cm 2 . The kinetics of ion thermosorption after corrosion of irradiated materials is studied, the temperature threshold of implanted ion stability in zirconium and its alloys is found to be 400 deg C

Method of fabricating zirconium metal for use in composite type fuel cans

International Nuclear Information System (INIS)

Imahashi, Hiromichi; Inagaki, Masatoshi; Akabori, Kimihiko; Tada, Naofumi; Yasuda, Tetsuro.

1985-01-01

Purpose: To mass produce zirconium metal for fuel cans with less radiation hardening. Method: Zirconium sponges as raw material are inserted in a hearth mold and a procedure of melting the zirconium sponges portionwise by using a melting furnace having electron beams as a heat source while moving the hearth is repeated at least for once. Then, the rod-like ingot after melting is melted again in a vacuum or inert gas atmosphere into an ingot of a low oxygen density capable of fabrication. A composite fuel can billet is formed by using the thus obtained zirconium ingot and a zircalloy, and a predetermined composite type fuel can is manufactured by way of hot extrusion and pipe drawing fabrication. The raw material usable herein is zirconium sponge with an oxygen density of 400 ppm or higher and the content of impurity other than oxygen is between 1000 - 5000 ppm in total, or the molten material thereof. (Kamimura, M.)

Mechanistic understanding of irradiation corrosion of zirconium alloys in nuclear power plants: stimuli, status and outlook

International Nuclear Information System (INIS)

Cox, B.; Ishigure, K.; Johnson, A.B.; Lemalgnan, J.C.; Nechaev, A.F.; Petrik, N.G.; Reznichenko, E.A.

1990-01-01

Extensive information about the corrosion behaviour of zirconium alloys under irradiation is presented. Review of the existing models of radiation corrosion is given. An accent is made on a necessity in conducting basic investigations to overcome contradictions in interpreting the experimental data available. Importance of solving the problem of zirconium alloy corrosion for safe NPP operation is underlined. 34 refs.; 6 figs.; 4 tabs

Strengthening and elongation mechanism of Lanthanum-doped Titanium-Zirconium-Molybdenum alloy

Energy Technology Data Exchange (ETDEWEB)

Hu, Ping, E-mail: huping1985@126.com [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Hu, Bo-liang; Wang, Kuai-she; Song, Rui; Yang, Fan [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Yu, Zhi-tao [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Tan, Jiang-fei [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Cao, Wei-cheng; Liu, Dong-xin; An, Geng [Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Guo, Lei [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Yu, Hai-liang [School of Mechanical, Materials and Mechatronics Engineering, University of Wollongong, NSW 2522 (Australia)

2016-12-15

The microstructural contributes to understand the strengthening and elongation mechanism in Lanthanum-doped Titanium-Zirconium-Molybdenum alloy. Lanthanum oxide particles not only act as heterogeneous nucleation core, but also act as the second phase to hinder the grain growth during sintering crystallization. The molybdenum substrate formed sub-grain under the effect of second phase when the alloy rolled to plate.

Improved nuclear fuel element

International Nuclear Information System (INIS)

1980-01-01

The invention is of a nuclear fuel element which comprises a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof, and an elongated composite cladding container comprising a zirconium alloy tube containing constituents other than zirconium in an amount greater than about 5000 parts per million by weight and an undeformed metal barrier of moderate purity zirconium bonded to the inside surface of the alloy tube. The container encloses the core so as to leave a gap between the container and the core during use in a nuclear reactor. The metal barrier is of moderate purity zirconium with an impurity level on a weight basis of at least 1000ppm and less than 5000ppm. Impurity levels of specific elements are given. Variations of the invention are also specified. The composite cladding reduces chemical interaction, minimizes localized stress and strain corrosion and reduces the likelihood of a splitting failure in the zirconium alloy tube. Other benefits are claimed. (U.K.)

Minimizing hydride cracking in zirconium alloys

International Nuclear Information System (INIS)

Coleman, C.E.; Cheadle, B.A.; Ambler, J.F.R.; Eadie, R.L.

1985-01-01

Zirconium alloy components can fail by hydride cracking if they contain large flaws and are highly stressed. If cracking in such components is suspected, crack growth can be minimized by following two simple operating rules: components should be heated up from at least 30K below any operating temperature above 450K, and when the component requires cooling to room temperature from a high temperature, any tensile stress should be reduced as much and as quickly as is practical during cooling. This paper describes the physical basis for these rules

Determination of zirconium by fluoride ion selective electrode

International Nuclear Information System (INIS)

Mahanty, B.N.; Sonar, V.R.; Gaikwad, R.; Raul, S.; Das, D.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

2010-01-01

Full text: Zirconium is used in a wide range of applications including nuclear clad, catalytic converters, surgical appliances, metallurgical furnaces, superconductors, ceramics, lamp filaments, anti corrosive alloys and photographical purposes. Irradiation testing of U-Zr and U-Pu-Zr fuel pins has also demonstrated their feasibility as fuel in liquid metal reactors. Different methods that are employed for the determination of zirconium are spectrophotometry, potentiometry, neutron activation analysis and mass spectrometry. Ion-selective electrode (ISE), selective to zirconium ion has been studied for the direct potentiometric measurements of zirconium ions in various samples. In the present work, an indirect method has been employed for the determination of zirconium in zirconium nitrate sample using fluoride ion selective electrode. This method is based on the addition of known excess amount of fluoride ion to react with the zirconium ion to produce zirconium tetra fluoride at about pH 2-3, followed by determination of residual fluoride ion selective electrode. The residual fluoride ion concentrations were determined from the electrode potential data using calibration plot. Subsequently, zirconium ion concentrations were determined from the concentration of consumed fluoride ions. A precision of about 2% (RSD) with the mean recovery of more than 94% has been achieved for the determination of zirconium at the concentration of 4.40 X 10 -3 moles lit -1

Nitrogen annealing of zirconium or titanium metals and their alloys

International Nuclear Information System (INIS)

Eucken, C.M.

1982-01-01

A method is described of continuously nitrogen annealing zirconium and titanium metals and their alloys at temperatures at from 525 0 to 875 0 C for from 1/2 minute to 15 minutes. The examples include the annealing of Zircaloy-4. (U.K.)

Microstructural changes in zirconium alloy bar due to multi-roll straightening

International Nuclear Information System (INIS)

Gouraharidas; Acharya, Swaroop; Pratap, Y.; Chaube, R.K.; Kiran Kumar, I.; Ramana Rao, A.V.; Saibaba, N.

2010-01-01

Zirconium alloy bar is the input material for making of end plugs required for encapsulating the uranium di-oxide pellets in the fuel tubes. These bars are manufactured through extrusion followed by multi-pass swaging with intermediate and final vacuum annealing. The straightened and ground bars are subjected to 100% Ultrasonic testing and Eddy current testing to identify flaws and micro-porosity in the material, which could otherwise affect the integrity of fuel element. The defect standards at ultrasonic and eddy current inspection have been made more stringent, in view of the importance of fuel pin integrity during reactor operation. Consequently, many of the rods have shown eddy current indications greater than the defect standard. Detailed microstructural examination was carried out at each process step to identify the cause for these indications. Characteristic variation in the grain size and microstructure were noticed from surface to the centre of the material. Correlation between residual stresses and the eddy current signals was established. The extent of residual stresses could be controlled by adopting improvised straightening method at the final stage. This paper deals with the various trials carried out and the conclusions arrived at. (author)

Effects of scandium and zirconium combination alloying on as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy

Directory of Open Access Journals (Sweden)

Xiang Qingchun

2011-02-01

Full Text Available The influences of minor scandium and zirconium combination alloying on the as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy have been experimentally investigated. The experimental results show that when the minor elements of scandium and zirconium are simultaneously added into the Al-4Cu-1.5Mg alloy, the as-cast microstructure of the alloy is effectively modified and the grains of the alloy are greatly refined. The coarse dendrites in the microstructure of the alloy without Sc and Zr additions are refined to the uniform and fine equiaxed grains. As the additions of Sc and Zr are 0.4% and 0.2%, respectively, the tensile strength, yield strength and elongation of the alloy are relatively better, which are 275.0 MPa, 176.0 MPa and 8.0% respectively. The tensile strength is increased by 55.3%, and the elongation is nearly raised three times, compared with those of the alloy without Sc and Zr additions.

An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys

International Nuclear Information System (INIS)

Fidleris, V.; Tucker, R.P.; Adamson, R.B.

1987-01-01

This paper presents an overview of factors affecting irradiation growth of zirconium alloys. Recent data obtained from irradiation programs in EBR-II, ATR, and NRU reactors are used to illustrate the effects of various microstructural and experimental factors on the growth of Zircaloy, zirconium, and zirconium-biobium alloys irradiated to fluences up to 2 X 10 26 nm -2 (E > 1 MeV) over the temperature range 330 to 720 K. Open literature results are also used to confirm or illustrate various effects. Important factors are texture, grain boundary parameters, residual stresses, original dislocation density, microstructure evolution, temperature during irradiation, solute effects, and fluence

Study of point defect clustering in electron and ion irradiated zirconium alloys

International Nuclear Information System (INIS)

Hellio, C.; Boulanger, L.

1986-09-01

Dislocation loops created by 500 keV Zr + ions and 1 MeV electrons in zirconium have a/3 type Burgers vectors, and in ion irradiated samples, loops lie preferentially on planes close to (1010). From in-situ observations of loop growth under 1 MeV electron irradiation in zirconium and dilute Zr (Nb,O) alloys, a strong increase of the vacancy migration energy with oxygen concentration was observed, from 0.72 eV for pure zirconium to 1.7 eV for Zr and Zr-1% Nb doped with 1800 ppm weight oxygen, indicating large trapping of vacancies by O single interstitials or clusters

Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

2001-02-01

Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

2001-02-01

Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily

Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

International Nuclear Information System (INIS)

Abraham, D.P.; McDeavitt, S.M.; Park, J.

1996-01-01

Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys

History of the development of zirconium alloys for use in nuclear reactors

International Nuclear Information System (INIS)

Rickover, H.G.; Geiger, L.D.; Lustman, B.

1975-01-01

The technical problems and the major decisions made during the early development of zirconium alloys for use in naval reactors are outlined. A summary is given of the development of commercial sources of supply for zirconium and hafnium metal over the period 1950 to 1965, and the problems encountered in obtaining zirconium needed for early naval prototype and shipboard reactors are identified. Steps taken in the Government procurement process are described and statistics on production amounts, prices, and inventory are included. Also included are the technical aspects associated with the development of zirconium for water-cooled nuclear reactors, beginning in early 1949 when the Bettis Atomic Power Laboratory was established as a part of the Naval Reactors Program. While in the course of the next 25 years, small-scale investigations were performed on other potential core structural materials such as stainless steel, niobium, aluminum, and beryllium, the pressure for continual development, improvement, and application of zirconium was predominant and unrelenting. (U.S.)

Oxidation kinetics and auger microprobe analysis of some oxidized zirconium alloys

International Nuclear Information System (INIS)

Ploc, R.A.

1989-01-01

Oxidation kinetics at 300 o C in dry oxygen of 0.5 wt% binary alloys of iron, nickel, and chromium in zirconium were determined for several surface preparations. Further, chemical profiles of the oxides as they existed on the matrix and on the precipitates were obtained by sputtering and Auger electron analysis. The appearance of 'breakaway' oxidation was controlled by the surface finish of the alloy, a variable that could be used to eliminate the phenomenon for all alloys except the Zr/Ni binary, which required β-quenching to accomplish the same purpose. (author)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

Energy Technology Data Exchange (ETDEWEB)

Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

International Nuclear Information System (INIS)

Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

2017-01-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Effects of alloying elements on nodular and uniform corrosion resistance of zirconium-based alloys

International Nuclear Information System (INIS)

Abe, Katsuhiro

1992-01-01

The effects of alloying and impurity elements (tin, iron, chromium, nickel, niobium, tantalum, oxygen, aluminum, carbon, nitrogen, silicon, and phosphorus) on the nodular and uniform corrosion resistance of zirconium-based alloys were studied. The improving effect of iron, nickel and niobium in nodular corrosion resistance were observed. The uniform corrosion resistance was also improved by nickel, niobium and tantalum. The effects of impurity elements, nitrogen, aluminum and phosphorus were negligibly small but increasing the silicon content seemed to improve slightly the uniform corrosion resistance. Hydrogen pick-up fraction were not changed by alloying and impurity elements except nickel. Nickel addition increased remarkably hydrogen pick-up fraction. Although the composition of secondary precipitates changed with contents of alloying elements, the correlation of composition of secondary precipitates to corrosion resistance was not observed. (author)

Experimental and numerical study of the effects of a nanocrystallisation treatment on high-temperature oxidation of a zirconium alloy

International Nuclear Information System (INIS)

Panicaud, B.; Retraint, D.; Grosseau-Poussard, J.-L.; Li, L.; Guérain, M.; Goudeau, P.; Tamura, N.; Kunz, M.

2012-01-01

Highlights: ► SMAT leads to a modification of surface properties of an M5 zirconium alloy (grain size and roughness. ► SMAT induces a change in the oxidation kinetics during high temperature oxidation. ► A diffusion model is able to reproduce kinetics and emphasise the consequences of SMAT on dissolution of oxygen in Zr. - Abstract: In the present work, the effects of a nanocrystallisation treatment on the high-temperature oxidation of a zirconium alloy are investigated. Surface Mechanical Attrition Treatment is a recent process designed to nanocrystallise the surface of materials. The particular effects of this treatment on an M5 zirconium alloy are studied using different experimental techniques at several scales. This material is of considerable interest, especially to the nuclear industry where very stringent conditions apply. High temperature oxidation was performed in order to show the benefits of this type of nanocrystallisation on the corrosion resistance of the alloy concerned. Microstructure development mechanisms, which improve the oxidation resistance of zirconium alloys have been identified during high-temperature corrosion. Those mechanisms have been discussed in further detail in relation to numerical calculations concerning the oxidation kinetics.

Electrochemical formation of uranium-zirconium alloy in LiCl-KCl melts

Energy Technology Data Exchange (ETDEWEB)

Murakami, Tsuyoshi, E-mail: m-tsuyo@criepi.denken.or.j [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Kato, Tetsuya; Kurata, Masaki [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Yamana, Hajimu [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

2009-11-15

Since zirconium is considered an electrochemically active species under practical conditions of the electrorefining process, it is crucial to understand the electrochemical behavior of zirconium in LiCl-KCl melts containing actinide ions. In this study, the electrochemical codeposition of uranium and zirconium on a solid cathode was performed. It was found that the delta-(U, Zr) phase, which is the only intermediate phase of the uranium-zirconium binary alloy system, was deposited on a tantalum substrate by potentiostatic electrolysis at -1.60 V (vs. Ag{sup +}/Ag) in LiCl-KCl melts containing 0.13 in mol% UCl{sub 3} and 0.23 in mol% ZrCl{sub 4} at 773 K. To our knowledge, this is the first report on the electrochemical formation of the delta-(U, Zr) phase. The relative partial molar properties of uranium in the delta-(U, Zr) phase were evaluated by measuring the open-circuit-potentials of the electrochemically prepared delta-phase electrode.

Electrochemical formation of uranium-zirconium alloy in LiCl-KCl melts

International Nuclear Information System (INIS)

Murakami, Tsuyoshi; Kato, Tetsuya; Kurata, Masaki; Yamana, Hajimu

2009-01-01

Since zirconium is considered an electrochemically active species under practical conditions of the electrorefining process, it is crucial to understand the electrochemical behavior of zirconium in LiCl-KCl melts containing actinide ions. In this study, the electrochemical codeposition of uranium and zirconium on a solid cathode was performed. It was found that the δ-(U, Zr) phase, which is the only intermediate phase of the uranium-zirconium binary alloy system, was deposited on a tantalum substrate by potentiostatic electrolysis at -1.60 V (vs. Ag + /Ag) in LiCl-KCl melts containing 0.13 in mol% UCl 3 and 0.23 in mol% ZrCl 4 at 773 K. To our knowledge, this is the first report on the electrochemical formation of the δ-(U, Zr) phase. The relative partial molar properties of uranium in the δ-(U, Zr) phase were evaluated by measuring the open-circuit-potentials of the electrochemically prepared δ-phase electrode.

Nuclear fuel elements

International Nuclear Information System (INIS)

Obara, Hiroshi.

1981-01-01

Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

Composition and Performance of Nanostructured Zirconium Titanium Conversion Coating on Aluminum-Magnesium Alloys

Directory of Open Access Journals (Sweden)

Sheng-xue Yu

2013-01-01

Full Text Available Nanostructured conversion coating of Al-Mg alloy was obtained via the surface treatment with zirconium titanium salt solution at 25°C for 10 min. The zirconium titanium salt solution is composed of tannic acid 1.00 g·L−1, K2ZrF6 0.75 g·L−1, NaF 1.25 g·L−1, MgSO4 1.0 g/L, and tetra-n-butyl titanate (TBT 0.08 g·L−1. X-ray diffraction (XRD, X-ray photoelectron spectroscopy (XPS, and Fourier transform infrared spectrum (FT-IR were used to characterize the composition and structure of the obtained conversion coating. The morphology of the conversion coating was obtained by atomic force microscopy (AFM and scanning electron microscopy (SEM. Results exhibit that the zirconium titanium salt conversion coating of Al-Mg alloy contains Ti, Zr, Al, F, O, Mg, C, Na, and so on. The conversion coating with nm level thickness is smooth, uniform, and compact. Corrosion resistance of conversion coating was evaluated in the 3.5 wt.% NaCl electrolyte through polarization curves and electrochemical impedance spectrum (EIS. Self-corrosion current density on the nanostructured conversion coating of Al-Mg alloy is 9.7×10-8A·cm-2, which is only 2% of that on the untreated aluminum-magnesium alloy. This result indicates that the corrosion resistance of the conversion coating is improved markedly after chemical conversion treatment.

Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M.; Howells, R. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

2014-11-15

Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal–oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.

Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

Science.gov (United States)

Platt, P.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2014-11-01

Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal-oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.

Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

International Nuclear Information System (INIS)

Platt, P.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2014-01-01

Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal–oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations

Influence of alloying elements on the dislocation loops created by Zr+ ion irradiation in alpha-zirconium

International Nuclear Information System (INIS)

Hellio, C.; Novion, C.H. de; Boulanger, L.

1987-01-01

Pure zirconium and four (annealed) α - zirconium based alloys (Zr-1760 ppm weight 0, Zr - 1% Nb - 430 ppm 0, Zr-1% Nb-1800 ppm 0, zircaloy 4) have been studied by transmission electron microscopy after 500 keV Zr + ion or 1 MeV electron irradiation performed at high temperature. Type of burgers vectors of the dislocation loops are given; in the case of electron irradiated Zr-1760 ppm 0, the larger loops were found of interstitial type. Alloying elements increase the loop density. The kinetic of loop growth was observed in-situ during 1 MeV electron irradiation between 400 and 700 0 C: oxygen was found to reduce considerably the growth speed of loops. In-situ annealing at 450 or 500 0 C after ion irradiation led to a large coalescence of loops in the case of pure zirconium, but modified only slightly the defect structure of the alloys

Enhanced low-temperature oxidation of zirconium alloys under irradiation

International Nuclear Information System (INIS)

Cox, B.; Fidleris, V.

1989-01-01

The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

High density fuels using dispersion and monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-07-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

High density fuels using dispersion and monolithic fuel

International Nuclear Information System (INIS)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

2017-01-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Onimus, F.

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Influence of hydratation on the characteristics of zirconium alloys oxide layers

Czech Academy of Sciences Publication Activity Database

Gosmanová, G.; Kraus, I.; Kolega, M.; Vrtílková, V.; Weishauptová, Zuzana

2008-01-01

Roč. 54, č. 1 (2008), s. 1576-1580 ISSN 1210-0471 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : zirconium alloys * corrosion layer * hydrated ZrO2 Subject RIV: JF - Nuclear Energetics

Influence de l’irradiation et de la radiolyse sur la vitesse et les mécanismes de corrosion des alliages de zirconium

OpenAIRE

Verlet , Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO2 pe...

Recycling melting process of the zirconium alloy chips

Energy Technology Data Exchange (ETDEWEB)

Reis, Luis A.M. dos; Mucsi, Cristiano S.; Tavares, Luiz A.P.; Alencar, Maicon C.; Gomes, Maurilio P.; Barbosa, Luzinete P.; Rossi, Jesualdo L., E-mail: luisreis.09@gmail.com, E-mail: csmucsi@gmail.com [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Pressurized water reactors (PWR) commonly use {sup 235}U enriched uranium dioxide pellets as a nuclear fuel, these are assembled and stacked in zirconium alloy tubes and end caps (M5, Zirlo, Zircaloy). During the machining of these components large amounts of chips are generated which are contaminated with cutting fluid. Its storage presents safety and environmental risks due to its pyrophoric and reactive nature. Recycling industry shown interest in its recycling due to its strategic importance. This paper presents a study on the recycling process and the results aiming the efficiency in the cleaning process; the quality control; the obtaining of the pressed electrodes and finally the melting in a Vacuum Arc Remelting furnace (VAR). The recycling process begins with magnetic separation of possible ferrous alloys chips contaminant, the washing of the cutting fluid that is soluble in water, washing with an industrial degreaser, followed by a rinse with continuous flow of water under high pressure and drying with hot air. The first evaluation of the process was done by an Energy Dispersive X-rays Fluorescence Spectrometry (EDXRFS) showed the presence of 10 wt. % to 17 wt. % of impurities due the mixing with stainless steel machining chips. The chips were then pressed in a custom-made matrix of square section (40 x 40 mm - 500 mm in length), resulting in electrodes with 20% of apparent density of the original alloy. The electrode was then melted in a laboratory scale VAR furnace at the CCTM-IPEN, producing a massive ingot with 0.8 kg. It was observed that the samples obtained from Indústrias Nucleares do Brasil (INB) are supposed to be secondary scrap and it is suggested careful separation in the generation of this material. The melting of the chips is possible and feasible in a VAR furnace which reduces the storage volume by up to 40 times of this material, however, it is necessary to correct the composition of the alloy for the melting of these ingots. (author)

Recycling melting process of the zirconium alloy chips

International Nuclear Information System (INIS)

Reis, Luis A.M. dos; Mucsi, Cristiano S.; Tavares, Luiz A.P.; Alencar, Maicon C.; Gomes, Maurilio P.; Barbosa, Luzinete P.; Rossi, Jesualdo L.

2017-01-01

Pressurized water reactors (PWR) commonly use 235 U enriched uranium dioxide pellets as a nuclear fuel, these are assembled and stacked in zirconium alloy tubes and end caps (M5, Zirlo, Zircaloy). During the machining of these components large amounts of chips are generated which are contaminated with cutting fluid. Its storage presents safety and environmental risks due to its pyrophoric and reactive nature. Recycling industry shown interest in its recycling due to its strategic importance. This paper presents a study on the recycling process and the results aiming the efficiency in the cleaning process; the quality control; the obtaining of the pressed electrodes and finally the melting in a Vacuum Arc Remelting furnace (VAR). The recycling process begins with magnetic separation of possible ferrous alloys chips contaminant, the washing of the cutting fluid that is soluble in water, washing with an industrial degreaser, followed by a rinse with continuous flow of water under high pressure and drying with hot air. The first evaluation of the process was done by an Energy Dispersive X-rays Fluorescence Spectrometry (EDXRFS) showed the presence of 10 wt. % to 17 wt. % of impurities due the mixing with stainless steel machining chips. The chips were then pressed in a custom-made matrix of square section (40 x 40 mm - 500 mm in length), resulting in electrodes with 20% of apparent density of the original alloy. The electrode was then melted in a laboratory scale VAR furnace at the CCTM-IPEN, producing a massive ingot with 0.8 kg. It was observed that the samples obtained from Indústrias Nucleares do Brasil (INB) are supposed to be secondary scrap and it is suggested careful separation in the generation of this material. The melting of the chips is possible and feasible in a VAR furnace which reduces the storage volume by up to 40 times of this material, however, it is necessary to correct the composition of the alloy for the melting of these ingots. (author)

Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Energy Technology Data Exchange (ETDEWEB)

A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

2009-08-01

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

Microstructural aspects of the oxidation of zirconium alloys

International Nuclear Information System (INIS)

Proff, Ch.

2011-01-01

This thesis is focused on the microstructural characterisation of precipitates in the oxide of binary zirconium alloys (1 wt.% Fe, Cr or Ni or 0.6 wt.% Nb) under different oxidation conditions at 415 C. The samples were oxidised in autoclave in air and steam and in an environmental scanning electron microscope in water vapour. The microstructural evolution of the precipitates during oxidation was characterised using electron microscopy. The findings from the analysis are the following: -Two types of oxidation behaviour are observed for precipitates. -Pilling Bedworth ratio of precipitates is higher than that of the zirconium matrix. -Formation of pure iron oxide crystals on the surface for iron bearing precipitates close to or at the surface. From these observations it is concluded that the precipitate oxidation behaviour can be correlated to precipitate composition and oxidation tendency of the elements in the precipitates. Iron exhibits clearly different behaviour. (author)

Round robin test for zirconium alloys in 400 deg C steam: results from EDF; Essais interlaboratoires de corrosion generalisee en milieu vapeur a 400 deg C d`alliages de zirconium: resultats d`EDF

Energy Technology Data Exchange (ETDEWEB)

Blat, M.

1994-01-01

The EDF Material Studies Branch has participated in the Round Robin program of uniform corrosion on zirconium alloys. The objectives of these Round Robin corrosion tests are to generate new uniform corrosion weight gain date utilizing modern zirconium alloy products and to improve the International and ASTM standards. (author). 2 tabs., 7 appendix., 2 refs.

Dilatometric studies on uranium-zirconium-fissium alloy

International Nuclear Information System (INIS)

Banerjee, Aparna; Kulkarni, S.G.; Kulkarni, R.V.; Kaity, Santu

2012-01-01

The knowledge of thermophysical properties of U-Zr alloys are important for modelling fuel behaviour in nuclear reactor. Fissium is an alloy that approximates the equilibrium concentration of the metallic fission product elements left by metallurgical reprocessing. Coefficient of thermal expansion (CTE) data is needed to calculate stresses occurring in fuel and cladding with change in temperature. Coefficient of thermal expansion can be utilized to determine the change of alloy density as a function of temperature. In the present investigation, thermophysical properties like coefficient of thermal expansion and density were determined using dilatometer for U-20wt.%Zr-5wt.%Fs alloy prepared by arc melting process. The microstructural investigation was carried out using scanning electron microscope

Nuclear fuel element

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

Components made of corrosion-resistent zirconium alloy and method for its production

International Nuclear Information System (INIS)

Hanneman, R.E.; Urquhart, A.W.; Vermilyea, D.A.

1977-01-01

The invention deals with a method to increase the resistance of zirconium alloys to blister corrosion which mainly occurs in boiling-water nuclear reactors. According to the method described, the surface of the alloy body is coated with a thin film of a suitable electronically conducting material. Gold, silver, platinum, nickel, chromium, iron and niobium are suitable as coating materials. The invention is more closely explained by means of examples. (GSC) [de

Preparation and certification of certified reference materials JAERI-Z21, Z22 and Z23 for analysis of zirconium and its alloys

International Nuclear Information System (INIS)

Takashima, Kyoichiro

1991-03-01

The Sub-Committee on Chemical Analysis of Nuclear Materials was organized in April 1987, under the Committee on Analytical Chemistry of Nuclear Fuels and Reactor Materials, JAERI, for renewal of certified reference materials of zirconium base alloys and zirconium metal. Collaborative analysis was carried out among ten participating laboratories for the certification of the JAERI CRMs Z21 to Z23. As a results of the collaborative works, the certified values for sixteen elements (Sn, Fe, Ni, Cr, Hf, Al, Si, Co, Cu, Ti, Mn, Pb, U, Cd, B and W) in the CRMs were given. In this report, preparation of raw materials, homogeneity test, chemical analysis for certification by collaborative works during April 1987 to March 1990 are described. (author)

Characterization of zirconium alloy oxidation films by alternating current impedance

International Nuclear Information System (INIS)

Rosecrans, P.M.

1983-11-01

Kinetics of zirocnium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Phase composition and properties of rapidly cooled aluminium-zirconium-chromium alloys

International Nuclear Information System (INIS)

Sokolovskaya, E.M.; Badalova, L.M.; Podd''yakova, E.I.; Kazakova, E.F.; Loboda, T.P.; Gribanov, A.V.

1989-01-01

Using the methods of physicochemical analysis the interaction of aluminium with zirconium and chromium is studied. Polythermal cross sections between Al 3 -Zr-Al 7 Cr and radial polythermal cross section from aluminium-rich corner with the ratio of components Zr:Cr=5:7 by mass are constructed. The effect of zirconium and chromium content on electrochemical characteristics of aluminium-base rapidly quenching alloys in systems Al-Cr, Al-Zr, Al-Cr-Zr. An increase in chromium concentration in oversaturated solid solution of Al-Cr system expands considerably the range of passive state. When Al 7 Cr phase appears the range of passive stae vanishes

Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

2003-05-01

Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

Fluorimetric determination of uranium in zirconium and zircaloy alloys; Determinacion fluorimetrica de uranio en aleaciones de zirconio y zircaloy

Energy Technology Data Exchange (ETDEWEB)

Acosta L, E [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1991-05-15

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

Growth and characterization of oxide layers on zirconium alloys

International Nuclear Information System (INIS)

Maroto, A.J.G.; Bordoni, R.; Villegas, M.; Blesa, M.A.; Olmedo, A.M.; Iglesias, A.; Rigotti, G.

1997-01-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab

Growth and characterization of oxide layers on zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Maroto, A J.G.; Bordoni, R; Villegas, M; Blesa, M A; Olmedo, A M; Iglesias, A; Rigotti, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1997-02-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab.

Innovative approaches in the manufacture of zirconium alloy components for PHWRs

International Nuclear Information System (INIS)

Rao, M.N.; Srivastava, R.K.

2005-01-01

Selection of an appropriate route for the fabrication of Zirconium alloy fuel components has a direct bearing on the quality of finished product. Many sophisticated and intricate processes such as vacuum arc melting, extrusion, hot rolling and cold working processes - swaging, drawing and sheet rolling are employed. Many advances were made in eddy current and ultrasonic evaluation to meet the stringent quality control requirement and locate the micro flaws. Emphasis was laid on achieving high recoveries and manufacture the product at minimum cost. Several creative and innovative processes were adopted particularly in the fabrication of end caps and spacers. The spacers were produced through the wire route and subsequently parting them into tiny spacers, which is entirely different from the conventional route of fabricating the sheets followed by blanking and coining. This has improved the material recovery and the lead time has been reduced substantially. The end caps used for the closure of clad tubes have to meet the most stringent quality requirements to avoid micro-flaws. The manufacturing processes adopted have direct influence on the integrity of the finished product. Special defect standards were developed to identify and eliminate micro-flaws and thereby ensure consistent and repetitive quality product. The paper brings out the above innovative approaches made in fabrication and quality control techniques in the manufacture of fuel components for PHWR fuel bundles. (author)

Hot-rolled and cold-finished zirconium and zirconium alloy bars, rod, and wire for nuclear application

International Nuclear Information System (INIS)

Anon.

1981-01-01

The specification covers hot- and cold-finished zirconium alloy bars, rod, and wire, other than those required for reforging, including rounds, squares, and shapes. One unalloyed grade and three alloy grades for use in nuclear applications are described. The products covered include the following sections and sizes: bars, rounds in coils for subsequent reworking (6.4 to 19 mm) and flats (6.4 to 250 mm); rods, rounds in coils for subsequent reworking (6.4 to 19 mm); wire (9.5 mm). The specification covers ordering information, manufacture, condition, chemical requirements, mechanical properties, corrosion properties, permissible variations in dimensions, significance of numerical limits, lot size, special tests, workmanship, finish, inspection, certification, packaging and marking

ZIRCONIUM-CLADDING OF THORIUM

Science.gov (United States)

Beaver, R.J.

1961-11-21

A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

Study of Influence of an Annealing on Corrosion Stability of Pipes-shells for Fuel of Zr1Nb Alloy

International Nuclear Information System (INIS)

Petel'guzov, I.A.; Rodak, A.G.; Pasenov, F.A.; Ishchenko, N.I.

2006-01-01

Explored influence an annealing to the kinetics of corrosion and mechanical characteristics of pipe material for shells fuel elements made from the experimental zirconium alloy Zr1Nb calcium-thermal way of production, in the comparison with the staff alloy E110 electrolytic way of production. Determined parameters of kinetics of corrosion depending on temperature and duration annealing before testing. Conducted also mechanical testing the alloys on the ring samples. Determined ranges of temperatures, within which corrosion characteristics save values, close to source, and connecting temperatures, under which is observed reduction research; investigating features

Mechanisms of irradiation growth of alpha-zirconium alloys

International Nuclear Information System (INIS)

Holt, R.A.

1988-01-01

Experimental observations in the last few years have shown that the range of irradiation growth behaviour of alpha-zirconium alloys is more varied, that a wider variety of sinks must be considered, and that there are more potential sources of anisotropy than was previously recognized. The important new experimental observations which influence our preception of the growth phenomenon in zirconium alloys include the growth of single crystals, accelerating growth in annealed material with the coincident appearance of vacancy loops on the basal planes, the occurrence of 'negative' growth, i.e., contractions along prism directions, the absence of a pronounced effect of grain size on the long term growth rate at low temperatures, and the presence of intergranular constraints prior to irradiation. With the greater complexity of behaviour now being observed, it is necessary to apply new theoretical concepts to assist in understanding growth, e.g., the potential role of anisotropic diffusion in segregation point defects to different sinks and 'growth' caused by the anisotropic relaxation of intergranular constrains. These can be combined with earlier ideas to predict a variety of growth behaviours, including 'negative growth'. Because the most important physical information required for theoretical treatments of growth, i.e, the characteristics of vacancies and self interstitial atoms, are still poorly understood, it is almost impossible to test rigorously any particular theoretical concept and a complete picture of growth has yet to emerge. (orig./MM)

Process for forming seamless tubing of zirconium or titanium alloys from welded precursors

International Nuclear Information System (INIS)

Sabol, G.P.; Barry, R.F.

1987-01-01

A process is described for forming seamless tubing of a material selected from zirconium, zirconium alloys, titanium, and titanium alloys, from welded precursor tubing of the material, having a heterogeneous structure resulting from the welding thereof. The process consists of: heating successive axial segments of the welded tubing, completely through the wall thereof, including the weld, to uniformly transform the heterogeneous, as welded, material into the beta phase; quenching the beta phase tubing segments, the heating and quenching effected sufficiently rapid enough to produce a fine sized beta grain structure completely throughout the precursor tubing, including the weld, and to prevent growth of beta grains within the material larger than 200 micrometers in diameter; and subsequently uniformly deforming the quenched precursor tubing by cold reduction steps to produce a seamless tubing of final size and shape

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

Energy Technology Data Exchange (ETDEWEB)

Onimus, F

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Determination of impurities in uranium--niobium (7.5%)--zirconium (2.5%) alloy

Energy Technology Data Exchange (ETDEWEB)

Arragon, Y

1973-10-01

The determination of 11 impurities in uranium--niobium-- zirconium alloys was studied. Elements of which the alloy is composed are considered and information is given on the determination of niobium by niobic acid precipitation. Selective elimination of the three components is discussed. Two liquid-liquid extractions are used. The nioblum is separated by methylisobutylketone in a hydrochloric --hydrofluoric medium and the zirconium and uranium by tributyl phosphate in a nitric medium. The determination of trace elements using electrochemical methods is discussed. Anodic re-dissolution polarography or square wave polarography enabled six elements (cadmium, copper, lead, zinc, bismuth, and thallium) to be determined in a carbonate medium together with aluminium in tetraethylammonium perchlorate, molybdenum in nitric acid, ammonium nitrate, and tungsten in hydrochloric acid with added double sodium and potassium tartrate. Fluorine was determined using ionometric techniques with a specific electrode and carbon was titrated by conductometry after combustion of the sample in an oxygen current. (auth)

Prevention of delayed hydride cracking in zirconium alloys

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.; Ambler, J.F.R.

1987-01-01

Zirconium alloys are susceptible to a mechanism for crack initiation and propagation called delayed hydride cracking. From a review of component failures and experimental results, we have developed the requirements for preventing this cracking. The important parameters for cracking are hydrogen concentration, flaws, and stress; each should be minimized. At the design and construction stages hydrogen pickup has to be controlled, quality assurance needs to be at a high enough level to ensure the absence of flaws, and residual stresses must be eliminated by careful fabrication and heat treatment

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

Energy Technology Data Exchange (ETDEWEB)

Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

2017-06-15

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

International Nuclear Information System (INIS)

Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

2017-01-01

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Study of the aqueous chemical treatment of uranium zirconium fuels; Etude du traitement chimique des combustibles uraniumzirconium par voie seche

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, M; Nollet, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

A dry process has been studied for separating the uranium from the zirconium-either for recovering the enriched uranium from fuel element production waste, or with a view to treating this waste after irradiation. In this process the alloy is treated with hydrochloric acid at 400 deg. C in a fluidized corundum bed which causes the zirconium to volatilize as tetrachloride and the uranium to form the trichloride. This latter is then converted to the hexafluoride by attack with fluorure. After the laboratory tests, a first pilot plant with a capacity of 1 kg of alloy was tried out at the Fontenay-aux-Roses Nuclear Research Centre; this made it possible to fix the operational conditions for the process. An industrial scale plant was then built with the collaboration of the from Kuhlmann, and operated until a satisfactory process had been developed for treating the waste. This installation treats 3 kg/h of alloy with a yield for the hydrochloric acid of about 50 per cent and with a uranium loss in the zirconium tetrachloride of about 0.1 per cent. An active pilot plant capable of treating of treating a few kilos of irradiated alloy is now being studied. (authors) [French] On a etudie un procede de voie seche pour effectuer la separation de l'uranium et du zirconium - soit en vue de la recuperation de l'uranium enrichi contenu dans les dechets de fabrication des elements combustibles - soit en vue du traitement de ceux-ci apres irradiation. Ce procede consiste a attaquer l'alliage par l'acide chlorhydrique a 400 deg. C dans un lit fluidise de corindon, ce qui a pour effet de volatiliser le zirconium sous forme de tetrachlorure et de transformer l'uranium en trichlorure. Ce dernier est ensuite converti en hexafluorure par action du fluor. Apres des essais de laboratoire, un premier pilote a l'echelle de 1 kg d'alliage a ete experimente au Centre d'Etudes Nucleaires de Fontenay-aux-Roses et a permis de determiner les conditions operatoires du procede. En collaboration avec

Method of processing spent fuel cladding tubes

International Nuclear Information System (INIS)

Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

1986-01-01

Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

Energy Technology Data Exchange (ETDEWEB)

Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

2007-07-01

Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Unloading Effect on Delayed Hydride Cracking in Zirconium Alloys

International Nuclear Information System (INIS)

Kim, Young Suk; Kim, Sung Soo

2010-01-01

It is well-known that a tensile overload retards not only the crack growth rate (CGR) in zirconium alloys during the delayed hydride cracking (DHC) tests but also the fatigue crack growth rate in metals, the cause of which is unclear to date. A considerable decrease in the fatigue crack growth rate due to overload is suggested to occur due either to the crack closure or to compressive stresses or strains arising from unloading of the overload. However, the role of the crack closure or the compressive stress in the crack growth rate remains yet to be understood because of incomplete understanding of crack growth kinetics. The aim of this study is to resolve the effect of unloading on the CGR of zirconium alloys, which comes in last among the unresolved issues as listed above. To this end, the CGRs of the Zr-2.5Nb tubes were determined at a constant temperature under the cyclic load with the load ratio, R changing from 0.13 to 0.66 where the extent of unloading became higher at the lower R. More direct evidence for the effect of unloading after an overload is provided using Simpson's experiment investigating the effect on the CGR of a Zr-2.5Nb tube of the stress states of the prefatigue crack tip by unloading or annealing after the formation of a pre-fatigue crack

Anelastic relaxation peaks in single crystals of zirconium-oxygen alloys

International Nuclear Information System (INIS)

Ritchie, I.G.; Sprungmann, K.W.; Atrens, A.; Rosinger, H.E.; CEA Centre d'Etudes Nucleaires de Grenoble, 38

1977-01-01

Relaxations of the compliances S 11 -S 12 and S 44 have been observed in single crystals of zirconium-oxygen alloys tested in flexure and in torsion respectively. The relaxations are attributed to the stress-induced reorientation of substitutional impurity atoms (s) paired with interstitial oxygen atoms (i). The results demonstrate that the jump of the interstitial parallel to the basal plane dominates in the reorientation of the s-i pair

Experimental study of the zirconium alloy oxidation under high pressure of steam and modelling of the mechanisms

International Nuclear Information System (INIS)

Dali, Yacoub

2007-01-01

The corrosion of the cladding materials used for the fuel rods is one of the limiting factor of their lifetime in light water reactors. In this field, the aim of the nuclear industry is today to increase the time and the number of cycles and to submit the claddings in zirconium alloys to higher corrosive conditions. In this way, new alloys devoted to replace the standard Zircaloy-4, for instance Nb containing alloys, have been recently developed and licensed and show better corrosion resistance. A better understanding of the corrosion mechanisms of the zirconium alloys is necessary to predict the corrosion behaviour of these materials. In this work, the oxidation rate of model alloys of two metallurgic families has been studied in steam in a pressure range between 100 milli-bars and 100 bars. The Zircaloy type alloys contain as alloying elements oxygen and/or tin and/or iron and chromium. For the Zr-Nb family, three niobium contents have been studied, respectively 0.2, 0.4 and 1 weight percent of niobium. Our objectives were to understand the variations of the reactivity between the low pressure and the high pressure range, in quantifying the dependency of the corrosion rate with the steam pressure and the alloying element concentrations. The segregation process of the niobium at the surface has also been studied on the Zr-Nb alloys. During this work, a magnetic suspension thermo-balance has been developed and used to follow in-situ the corrosion rate at high pressure of water vapour. The oxide layers have been characterized by many techniques, macro and micro-photo-electrochemistry, XRD, FEG-SEM, XPS, HR-TEM and SIMS. For the Zircaloy type alloys, we have confirmed the major role of the intermetallic precipitates Zr(Fe,Cr) 2 on the corrosion resistance. Unlike the standard Zircaloy-4, for which the oxidation rate does not depend on the pressure of the water vapour and is thus limited by the vacancy diffusion in the oxide layer, we have shown that the rate of the

Activation analysis in zirconium and alloys for nuclear application

International Nuclear Information System (INIS)

Cohen, I.M.; Mila, M.I.; Gomez, C.D.

1981-01-01

A study has been performed with the purpose to ascertain the possibilities of using neutron activation analysis in non-destructive determination of several elements present in zirconium and its alloys. Those elements must be limited within acceptable top levels, in accordance to standards for nuclear applications. The experimental techniques used are described and the results obtained are discussed, showing that the method is adequate for determining Cl, Co, Hf, Mn, and W, but not Ni and U. (M.E.L.) [es

Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

International Nuclear Information System (INIS)

Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

1996-01-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

Improved nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

International Nuclear Information System (INIS)

Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

1995-01-01

Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

A comparative study on the fretting wear properties of advanced zirconium fuel cladding materials

International Nuclear Information System (INIS)

Lee, Young Ho; Kim, Hyung Kyu; Park, Jeong Yong; Kim, Jun Hwan

2005-06-01

Fretting wear tests were carried out in room and high temperature water in order to evaluate the wear properties of new zirconium nuclear fuel claddings (K2∼K6) and the commercial claddings (M5, zirlo and zircaloy-4). The objective is to compare the wear resistance of K2∼K6 claddings with that of the commercial ones at the same test condition. After the wear tests, the average wear volume and the maximum wear depth were evaluated and compared at each test condition. As a result, it is difficult to select the most wear-resistant cladding between the K2∼K6 claddings and the commercial ones. This is because the average wear volume and maximum depth of each cladding included between the scattering range of measured results. However, wear resistance of the tested claddings based on the average wear volume and maximum wear depth could be summarized as follows: K5 > zircaloy-4 > (K2,K3) > (K4,M5) > K6 > zirlo at room temperature, zircaloy-4 > K5 > (K3,K4,zirlo) > (K2,K6) > M5 at high temperature and pressure. Therefore, it is concluded that K5 cladding among the tested new zirconium alloys has relatively higher wear-resistance in room and high temperature condition. In order to examine the wear mechanism, it is necessary to systematically study with the consideration of the alloying element effect and test environment. In this report, the wear test procedure and the wear evaluation method are described in detail

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

International Nuclear Information System (INIS)

Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

2015-01-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

Influence of protecting gel film on oxidation of zirconium alloys

Czech Academy of Sciences Publication Activity Database

Frank, H.; Weishauptová, Zuzana; Vrtílková, V.

2007-01-01

Roč. 360, č. 3 (2007), s. 282-292 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : fuel cladding * corrosion * Zirconium oxide Subject RIV: JF - Nuclear Energetics Impact factor: 1.643, year: 2007

Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

International Nuclear Information System (INIS)

Megata, Masao

1997-01-01

A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

Spectrophotometric determination of zirconium in nickel-base alloys with Arsenazo III after separation by froth flotation

International Nuclear Information System (INIS)

Sekine, K.; Onishi, H.

1977-01-01

0.02-0.1% of zirconium can be determined in nickel alloys by spectrophotometry with Arsenazo III after its separation from the sample solution by means of froth flotation using Arsenazo III and Zephiramine. Nickel, chromium and iron do not interfere. Analysis of standard alloys yielded a standard deviation of 2.2%. (orig.) [de

Recent advancements of chemical engineering in front end fuel cycle technologies at NFC. Contributed Paper IT-01

International Nuclear Information System (INIS)

Saibaba, N.

2014-01-01

On front end fuel cycle side, Nuclear Fuel Complex (NFC) has been a pioneer in processing the uranium and zirconium ore concentrates from different sources. The uranium and zirconium ore concentrates are converted into nuclear grade uranium and zirconium di oxide powders through the conventional TBP purification and precipitation route. In case of zirconium powders, they are converted into pure nuclear grade zirconium sponge through chlorination route for the production of zirconium alloys, which are mainly used as reactor core structural material

Multiscale modelling of hydrogen embrittlement in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Majevadia, Jassel; Wenman, Mark; Balint, Daniel; Sutton, Adrian [Imperial College London (United Kingdom); Nazarov, Roman [MPIE, Dusseldorf (Germany)

2013-07-01

Delayed Hydride Cracking (DHC) is a commonly occurring embrittlement phenomenon in zirconium alloy fuel cladding within Pressurized Water Reactors (PWRs). DHC is caused by the accumulation of hydrogen atoms taken up by the metal, and the formation of brittle hydrides in the vicinity of crack tips. The rate of crack growth is limited by the rate of hydrogen diffusion to the crack, which can be modelled by solving a stress driven diffusion equation that incorporates the elastic interaction between defects. This of interest in the present work. The elastic interaction is calculated by combining defect forces determined through Density Functional Theory (DFT) simulations, and an exact solution for the anisotropic elastic field of an edge dislocation in Zr. making it possible to determine the interaction energy without the need to simulate directly a hydrogen atom in the presence of a crack or dislocation, which is computationally prohibitive with DFT. The result of the elastic interaction energy calculations can be utilised to determine the segregation of hydrogen to a crack tip for varying crack tip geometries, and in the presence of other crystal defects. This is done by implementing a diffusion equation for hydrogen within a discrete dislocation dynamics simulation. In the present work a model has been developed to demonstrate the effect of a single dislocation on hydrogen diffusion to create a Cottrell atmosphere.

Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

International Nuclear Information System (INIS)

Doan, D.T.

2009-01-01

In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

Modelling of Zirconium and Hafnium separation using continuous annular chromatography

International Nuclear Information System (INIS)

Moch-Setyadji; Endang Susiantini

2014-01-01

Nuclear degrees of zirconium in the form of a metal alloy is the main material for fuel cladding of NPP. Zirconium is also used as sheathing UO 2 kernel in the form of ZrC as a substitute of SiC in the fuel elements of High Temperature Reactor (HTR). Difficulty separating hafnium from zirconium because it has a lot of similarities in the chemical properties of Zr and Hf. Annular chromatography is a device that can be used for separating of zirconium and hafnium to obtain zirconium nuclear grade. Therefore, it is necessary to construct the mathematical modelling that can describe the separation of zirconium and hafnium in the annular chromatography containing anion resin dowex-1X8. The aim of research is to perform separation simulation by using the equilibrium model and mass transfer coefficient resulted from research. Zr and Hf feed used in this research were 26 and 1 g/l, respectively. Height of resin (L), angular velocity (ω) and the superficial flow rate (uz) was varied to determine the effect of each parameter on the separation of Zr and Hf. By using Kd and Dv values resulted previous research. Simulation results showed that zirconium and hafnium can be separated using a continuous annular chromatography with high resin (long bed) 50 cm, superficial flow rate of 0.001 cm/s, the rotation speed of 0.006 rad/min and 20 cm diameter annular. In these conditions the results obtained zirconium concentration of 10,303.226 g/m 3 and hafnium concentration of 12.324 g/m 3 (ppm). (author)

Evaluation of a Zirconium Recycle Scrubber System

Energy Technology Data Exchange (ETDEWEB)

Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

Energy Technology Data Exchange (ETDEWEB)

Coffey, Greg W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meinhardt, Kerry D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pederson, Larry R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that

Effects of titanium and zirconium on iron aluminide weldments

Energy Technology Data Exchange (ETDEWEB)

Mulac, B.L.; Edwards, G.R. [Colorado School of Mines, Golden, CO (United States). Center for Welding, Joining, and Coatings Research; Burt, R.P. [Alumax Technical Center, Golden, CO (United States); David, S.A. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1997-12-01

When gas-tungsten arc welded, iron aluminides form a coarse fusion zone microstructure which is susceptible to hydrogen embrittlement. Titanium inoculation effectively refined the fusion zone microstructure in iron aluminide weldments, but the inoculated weldments had a reduced fracture strength despite the presence of a finer microstructure. The weldments fractured by transgranular cleavage which nucleated at cracked second phase particles. With titanium inoculation, second phase particles in the fusion zone changed shape and also became more concentrated at the grain boundaries, which increased the particle spacing in the fusion zone. The observed decrease in fracture strength with titanium inoculation was attributed to increased spacing of second phase particles in the fusion zone. Current research has focused on the weldability of zirconium- and carbon-alloyed iron aluminides. Preliminary work performed at Oak Ridge National Laboratory has shown that zirconium and carbon additions affect the weldability of the alloy as well as the mechanical properties and fracture behavior of the weldments. A sigmajig hot cracking test apparatus has been constructed and tested at Colorado School of Mines. Preliminary characterization of hot cracking of three zirconium- and carbon-alloyed iron aluminides, each containing a different total concentration of zirconium at a constant zirconium/carbon ratio of ten, is in progress. Future testing will include low zirconium alloys at zirconium/carbon ratios of five and one, as well as high zirconium alloys (1.5 to 2.0 atomic percent) at zirconium/carbon ratios of ten to forty.

The outcome of the working visits by the experts to the fuel provider enterprises

International Nuclear Information System (INIS)

Mironov, Y.; Molchanov, V.

2015-01-01

The working visits by the international experts to the fabrication plants of nuclear fuels and the zirconium alloy component facility took part in the framework of the „Zero failure level“ project. The purpose of these working visits was to determine whether there are any cause-effect relationships between the production of nuclear fuel and the fuel failures as well as to identify the trends and reasons for such failures. As an outcome of the analysis of the production processes at the fabrication plants of the nuclear fuel and the zirconium alloys component facility, the system root causes affecting the failures of the nuclear fuel were not identified

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Accident tolerant fuels for LWRs: A perspective

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J., E-mail: zinklesj@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2014-05-01

The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

Accident tolerant fuels for LWRs: A perspective

International Nuclear Information System (INIS)

Zinkle, S.J.; Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L.

2014-01-01

The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms

Zirconium ignition in exposed fuel channel

Energy Technology Data Exchange (ETDEWEB)

Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

2015-05-15

Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

Acoustic emission from zirconium alloys during mechanical and fracture testing

International Nuclear Information System (INIS)

Coleman, C.E.

1986-10-01

The application of acoustic emission during the mechanical and fracture testing of zirconium alloys is reviewed. Acoustic emission is successful in following delayed hydride cracking quantitatively. It is especially useful when great sensitivity is required. Application to fatigue, tensile deformation and stress corrosion cracking appears promising but requires more work to separate phenomena before it can be used quantitatively. This report is based on an invited review for the American Society of Non-Destructive Testing Handbook: Volume 5, Acoustic Emission Testing

Method to electrolytically precipitate metals onto zirconium objects

International Nuclear Information System (INIS)

Donaghy, R.E.

1978-01-01

Tubes and other formed bodies made of zirconium or zirconium alloys which serve to take up nuclear fuels, are plated by electrolytically depositing a metal film onto these in order to improve their mechanical and corrosion properties. The object is activated in a solution of ammonium bifluoride and sulphuric acid, whereby an electrically conducting solid and a loose layer is formed. This loose film is removed by using fluoboric acid or hydrofluoric silicic acid solution, ultrasonics, or strips of organic material (cotton, polyester, nylon). The plating of Cu, Ni, Cr is described in detail. The object is rinsed between the process steps with deionized water and finally degased at a temperature of 150-200 0 C. (IHOE) [de

Methods of studying oxide scales grown on zirconium alloys in autoclaves and in a PWR

International Nuclear Information System (INIS)

Blank, H.; Bart, G.; Thiele, H.

1992-01-01

The analysis of water-side corrosion of zirconium alloys has been a field of research for more than 25 years, but the details of the mechanisms involved still cannot be put into a coherent picture. Improved methods are required to establish the details of the microstructure of the oxide scales. A new approach has been made for a general analysis of oxide specimens from scales grown on the zirconium-based cladding alloys of PWR rods in order to analyse the morphology of these scales, the topography of the oxide/metal interface and the crystal structures close to this interface: a) Instead of using the conventional pickling solutions, the Zr-alloys are dissolved using a 'softer' solution (Br 2 in an organic solvent) in order to avoid damage to the oxide at the oxide/metal interface to be analysed by SEM (scanning electron microscopy). A second advantage of this method is easy etching of the grain structure of Zr-alloys for SEM analysis; b) By using the particular properties of the oxide scales, the corrosion-rate-determining innermost part of the oxide layer at the oxide/metal interface can be separated from the rest of the oxide scale and then analysed by SEM, STEM (scanning transmission electron microscopy), TEM (transmission electron microscopy) and electron diffraction after dissolution of the alloy. Examples are given from oxides grown on Zr-alloys in a pressurized water reactor and in autoclaves. (author) 8 figs., 3 tabs., 9 refs

Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

International Nuclear Information System (INIS)

Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

2006-01-01

The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty, cycles high burnup, boiling, aggressive chemistry) and to investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment

Voltammetric determination of zirconium using azo compounds

International Nuclear Information System (INIS)

Orshulyak, O.O.; Levitskaya, G.D.

2008-01-01

The optimum conditions for zirconium complexation with azo compounds are found. The applicability of Eriochrome Red B, Calcon, and Calcion to the voltammetric determination of zirconium, total Zr(IV) and Hf(IV), and Zr(IV) in the presence of Zn(II), Cu(II), Cd(II), Ni(II), or Ti(IV) is demonstrated. The developed procedures are used to determine zirconium in a terbium alloy and in an alloy for airplane wheel drums [ru

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Ultrasonic texture characterization of aluminum, zirconium and titanium alloys

International Nuclear Information System (INIS)

Anderson, A.J.

1997-01-01

This work attempts to show the feasibility of nondestructive characterization of non-ferrous alloys. Aluminum alloys have a small single crystal anisotropy which requires very precise ultrasonic velocity measurements for derivation of orientation distribution coefficients (ODCs); the precision in the ultrasonic velocity measurement required for aluminum alloys is much greater than is necessary for iron alloys or other alloys with a large single crystal anisotropy. To provide greater precision, some signal processing corrections need to be applied to account for the inherent, half-bandwidth offset in triggered pulses when using a zero-crossing technique for determining ultrasonic velocity. In addition, alloys with small single crystal anisotropy show a larger dependence on the single crystal elastic constants (SCECs) when predicting ODCs which require absolute velocity measurements. Attempts were made to independently determine these elastics constants in an effort to improve correlation between ultrasonically derived ODCs and diffraction derived ODCs. The greater precision required to accurately derive ODCs in aluminum alloys using ultrasonic nondestructive techniques is easily attainable. Ultrasonically derived ODCs show good correlation with derivations made by Bragg diffraction techniques, both neutron and X-ray. The best correlation was shown when relative velocity measurements could be used in the derivations of the ODCs. Calculation of ODCs in materials with hexagonal crystallites can also be done. Because of the crystallite symmetries, more information can be extracted using ultrasonic techniques, but at a cost of requiring more physical measurements. Some industries which use materials with hexagonal crystallites, e.g. zirconium alloys and titanium, have traditionally used texture parameters which provide some specialized measure of the texture. These texture parameters, called Kearns factors, can be directly related to ODCs

Nuclear fuel cladding material

International Nuclear Information System (INIS)

Nakahigashi, Shigeo.

1982-01-01

Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

International Nuclear Information System (INIS)

Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

1977-11-01

This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

Zirconium-barrier cladding attributes

International Nuclear Information System (INIS)

Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

1987-01-01

This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

In-reactor creep of zirconium alloys by thermal spikes

International Nuclear Information System (INIS)

Ibrahim, E.F.

1975-01-01

The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for 2 x 10 -11 s at greater than melting point, at 570K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests-in-reactor, if spike lifetimes are 2 to 2.5 x 10 -11 s. (Auth.)

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 24-204 (United States)

2011-07-15

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Science.gov (United States)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin

2011-01-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Estimation of zirconium in various process streams in molten salt electrorefining process

International Nuclear Information System (INIS)

Suganthi, S.; Vandarkuzhali, S.; Venkatesh, P.; Prabhakara Reddy, B.; Nagarajan, K.

2012-01-01

Molten salt electrorefining process is a non-aqueous pyrochemical process suitable for reprocessing spent metallic fuel. In this process the spent fuel is taken at the anode and the fuel elements are selectively electrotransported to a suitable cathode (either a solid steel cathode or liquid cadmium cathode) using molten LiCl-KCI as electrolyte. We have demonstrated electrorefining of UZr alloy at engineering scale level. 1 Kg U-6%Zr alloy was taken at the anode and pure uranium was recovered at a steel cathode using molten LiCIKCI-5%UCI 3 as electrolyte at 773 K. In this paper we present the method of dissolution, sample preparation and estimation of zirconium in various process streams in the electrorefining experiments carried out in our laboratory

A layman's guide to radiation-induced deformation processes in zirconium alloys

International Nuclear Information System (INIS)

Dutton, R.

1990-07-01

The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

Hydrogenation and high temperature oxidation of Zirconium claddings

International Nuclear Information System (INIS)

Novotny, T.; Perez-Feró, E.; Horváth, M.

2015-01-01

In the last few years a new series of experiments started for supporting the new LOCA criteria, considering the proposals of US NRC. The effects which can cause the embrittlement of VVER fuel claddings were reviewed and evaluated in the framework of the project. The purpose of the work was to determine how the fuel cladding’s hydrogen uptake under normal operating conditions, effect the behavior of the cladding under LOCA conditions. As a first step a gas system equipment with gas valves and pressure gauge was built, in which the zirconium alloy can absorb hydrogen under controlled conditions. In this apparatus E110 (produced by electrolytic method, currently used at Paks NPP) and E110G (produced by a new technology) alloys were hydrogenated to predetermined hydrogen contents. According the results of ring compression tests the E110G alloys lose their ductility above 3200 ppm hydrogen content. This limit can be applied to determine the ductile-brittle transition of the nuclear fuel claddings. After the hydrogenation, high temperature oxidation experiments were carried out on the E110G and E110 samples at 1000 °C and 1200 °C. 16 pieces of E110G and 8 samples of E110 with 300 ppm and 600 ppm hydrogen content were tested. The oxidation of the specimens was performed in steam, under isothermal conditions. Based on the ring compression tests load-displacement curves were recorded. The main objective of the compression tests was to determine the ductile-brittle transition. These results were compared to the results of our previous experiments where the samples did not contain hydrogen. The original claddings showed more ductile behavior than the samples with hydrogen content. The higher hydrogen content resulted in a more brittle mechanical behavior. However no significant difference was observed in the oxidation kinetics of the same cladding types with different hydrogen content. The experiments showed that the normal operating hydrogen uptake of the fuel claddings

Aqueous corrosion study on U-Zr alloy

International Nuclear Information System (INIS)

Pal, Titas; Venkatesan, V.; Kumar, Pradeep; Khan, K.B.; Kumar, Arun

2009-01-01

In low power or research reactor, U-Zr alloy is a potential candidate for dispersion fuel. Moreover, Zirconium has a low thermal-neutron cross section and uranium alloyed with Zr has excellent corrosion resistance and dimensional stability during thermal cycling. In the present study aqueous corrosion behavior of U-Zr alloy samples was studied in autoclave at 200 deg C temperature. Corrosion rate was determined from weight loss with time. (author)

Evaluation of a Ductility after High Temperature Oxidation with the Three-Point Bend Test in Zirconium Alloys

International Nuclear Information System (INIS)

Jung, Yang Il; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

2010-01-01

In a light water reactor, the fuel cladding play an important role of preventing leakage of radioactive materials into the coolant, and thus the mechanical integrity of the cladding should be guaranteed under the conditions of normal and transient operation. In the case of a loss of coolant accident (LOCA), the cladding is subjected to a high temperature oxidation which is finally quenched because of an emergency coolant reflooding into the core. In this situation, the current LOCA criteria consist of five separate requirements: i) peak cladding temperature, ii) maximum cladding oxidation, iii) maximum hydrogen generation, iv) coolable geometry, and v) long-term cooling. The claddings lose their ductility due to the microstructural phase transformation from beta to martensite alpha-prime. and hydrogen up-take after LOCA. Since the reduction in ductility can induce embrittlement of claddings, post-quench ductility is one of the major concerns in transient operation circumstances. For the analysis, usually ring compression test are performed on ring samples cut from the tube to examine the oxidized cladding ductility. However, the test would not be applicable to the platelet samples which are general form of a specimen for developing alloys. As a high burn-up fuel cladding materials, Zircaloys are being replaced by modern zirconium alloys such as ZIRLO, and M5. Korea has also developed a new fuel cladding material HANA (high performance alloy for nuclear application) by the Korea Atomic Energy Research Institute. Because of the different composition of the newer claddings in comparison with the conventional Zircaloy-4, the high temperature oxidation behavior and the ductility after the oxidation would be different, and the properties should be evaluated how much the newer claddings were improved

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

Science.gov (United States)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Near net shape processing of zirconium or hafnium metals and alloys

International Nuclear Information System (INIS)

Evans, S.C.

1992-01-01

This patent describes a process for producing a metal shape. It comprises: plasma arc melting a metal selected from zirconium, hafnium and alloys thereof comprising at least about 90 w/o of these metals to form a liquid pool; pouring the metal form the pool into a mold to form a near net shape; and reducing the metal from its near net shape to a final size while maintaining the metal temperature below the alpha-beta transition temperature throughout the size reducing step

Properties of zirconium carbide for nuclear fuel applications

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

2013-10-15

Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

Influence of alloying elements on the irradiation hardening and environmental sensitivity of zirconium alloys

International Nuclear Information System (INIS)

Pettersson, K.; Hallstadius, L.; Bergqvist, H.; Nylund, A.; Wikstroem, C.

1992-01-01

Ten different alloys of zirconium have been tested with regard to the effect of irradiation on their mechanical properties and their sensitivity to environmentally induced failure. Two different environments were used: iodine vapour and liquid cesium with an addition of 2% cadmium. The neutron dose was 10 21 n/cm 2 (E>1MeV) and the irradiation temperature was about 300 degrees C. All alloy additions increased the irradiation hardening. Especially notable was the large effect of titanium and tin on irradiation hardening. A limited amount of transmission electron microscopy was carried out in order to find an explanation to the effects. The testing in different environments showed that there is no clear correlation between environmental sensitivity and yield stress. For materials of similar yield stress an alloyed material tends to be more sensitive to environmental cracking than a material which only contains oxygen as an impurity. There also seems to be an effect of oxygen on the environmental cracking sensitivity. A material with 910 ppm oxygen was considerably more sensitive to cracking than a material with 470 ppm oxygen despite the fact that the yield stress values differed by only 90 MPa

Superficial effects during the activation of zirconium AB2 alloys

International Nuclear Information System (INIS)

Zerbino, J; Visitin, A; Triaca, W

2005-01-01

The activation of zirconium nickel alloys with and without the addition of chromium and titanium is investigated through electrochemical and optical techniques.These alloys show high hydrogen absorption capacity and are extensively used in metal hydride batteries.Recent investigations in aqueous 1 M KOH indicate oxide layer growth and occlusion of hydrogen species in the alloys during the application of different cathodic potential programmes currently used in the activation process.In this research several techniques such as voltammetry, ellipsometry, energy dispersive analysis of X-rays EDAX, and scanning electron microscopy SEM are applied on the polished massive alloy Zr 1 -xTi x , x=0.36 y 0.43, and Zr 1 -xTi x CrNi, x=0.1,0.2 y 0.4.Data analysis shows that the stability, compactness and structure of the passive layers are strongly dependent on the applied potential programme.The alloy activation depends on the formation of deepen crevices that remain after a new polishing. Microscopic observation shows increase in the crevices thickness after the cathodic sweep potential cycling, which produces fragmentation of the grains and oxide growth during the activation process.This indicates metal breaking and intergranular dissolution that take place together with oxide and hydride formation.In some cases the resultant crevice thickness is one or two orders higher than that of the superficial oxide growth indicating intergranular localised corrosion

Plasma sprayed and electrospark deposited zirconium metal diffusion barrier coatings

International Nuclear Information System (INIS)

Hollis, Kendall J.; Pena, Maria I.

2010-01-01

Zirconium metal coatings applied by plasma spraying and electrospark deposition (ESD) have been investigated for use as diffusion barrier coatings on low enrichment uranium fuel for research nuclear reactors. The coatings have been applied to both stainless steel as a surrogate and to simulated nuclear fuel uranium-molybdenum alloy substrates. Deposition parameter development accompanied by coating characterization has been performed. The structure of the plasma sprayed coating was shown to vary with transferred arc current during deposition. The structure of ESD coatings was shown to vary with the capacitance of the deposition equipment.

Phase transformations in intermetallic phases in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Filippov, V. P., E-mail: vpfilippov@mephi.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation); Kirichenko, V. G. [Kharkiv National Karazin University (Ukraine); Salomasov, V. A. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation); Khasanov, A. M. [University of North Carolina – Asheville, Chemistry Department (United States)

2017-11-15

Phase change was analyzed in intermetallic compounds of zirconium alloys (Zr – 1.03 at.% Fe; Zr – 0.51 at.% Fe; Zr – 0.51 at.% Fe – M(M = Nb, Sn). Mössbauer spectroscopy on {sup 57}Fe nuclei in backscattering geometry with the registration of the internal conversion electrons and XRD were used. Four types of iron bearing intermetallic compounds with Nb were detected. A relationship was found between the growth process of intermetallic inclusions and segregation of these phases. The growth kinetics of inclusions possibly is not controlled by bulk diffusion, and a lower value of the iron atom’s activation energy of migration can be attributed to the existence of enhanced diffusion paths and interface boundaries.

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

Energy Technology Data Exchange (ETDEWEB)

STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

Microstructure and age-hardening effects of aluminium alloys with additions of scandium and zirconium

Energy Technology Data Exchange (ETDEWEB)

Galun, R.; Mordike, B.L. [Inst. fuer Werkstoffkunde und Werkstofftechnik, Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany); Maiwald, T.; Smola, B. [Zentrum fuer Funktionswerkstoffe GmbH, Clausthal-Zellerfeld (Germany); Mergen, R.; Manner, M.; Uitz, W. [Miba Gleitlager GmbH, Laakirchen (Australia)

2004-12-01

The aim of the work presented in this report was to produce age-hardenable aluminium alloys containing scandium and zirconium by a casting process with similar cooling conditions like an industrial casting process. Microstructure, precipitation structure and age-hardening response of different alloys with up to 0.4 wt.% Sc and Zr were investigated. Age-hardening experiments from the as-cast condition without solution annealing showed a significant increase of hardness of about 100% for Sc-rich alloys and of 50% for Zr-rich alloys compared to the as-cast condition. TEM investigations revealed the formation of precipitates of ternary Al{sub 3}(Sc{sub x}Zr{sub 1-x}) phases with a cubic cP4 crystal structure. In addition to the strengthening effect, a high thermal stability especially of the precipitates in Zr-rich alloys up to 400 C let these alloys look very promising for high-temperature applications. (orig.)

Investigation of Zirconium Oxide Films in Different Dissolved Hydrogen Concentration

International Nuclear Information System (INIS)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun

2016-01-01

It has been reported that in pre-transition zirconium oxide, the volume fraction of tetragonal zirconium oxide increased near the oxide/metal (O/M) interface, and the sub-stoichiometric zirconium oxide layer was observed. The diffusion of oxygen ion through the oxide layer is the rate-limiting process during the pre-transition oxidation process, and this diffusion mainly occurs in the grain boundaries. The two layered oxide structure is formed in pre-transition oxide for the zirconium alloy in high-temperature water environment. It is known that the corrosion rate is related to the volume fraction of zirconium oxide and the pores in the oxides; therefore, the aim of this paper is to investigate the oxidation behavior in the pre-transition zirconium oxide in high-temperature water chemistry. In this study, in situ Raman and TEM analysis were conducted for investigating the phase transformation of zirconium alloy in primary water. From this study, the following conclusions are drawn: 1. The zirconium alloy was oxidized in primary water chemistry for 100 d, and Raman and TEM were measured after 30, 50, 80, and 100 d from start-up. 2. TEM and FFT analysis showed that the zirconium oxide mostly consisted of the monoclinic phase. The tetragonal zirconium oxide was just found near the O/M interface

Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

International Nuclear Information System (INIS)

Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

1995-01-01

The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

Oxide characterization and hydrogen behaviors of Zr-based alloys

International Nuclear Information System (INIS)

Kim, Y. S.; Kim, D. J.; Kwon, S. H.; Lee, H. S.; Oh, S. J.; Yim, B. J.; Son, S. B.; Yun, S. P.

2006-03-01

The work scope and contents of the research are as follows : basic properties of zirconium alloys, hydrogen pick-up mechanism of zirconium alloy, effects of hydride on the corrosion behaviors of zirconium alloys, estimation on stress of oxide layer in the zirconium alloy, microstructure and characteristic of oxide in pre-hydrided zirconium alloys

Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

Science.gov (United States)

Rebak, Raul B.

2017-12-01

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

Effect of Electric Voltage and Current of X-ray Chamber on the Element inthe Zirconium Alloy Analysis X-ray by X-ray Fluorescence

International Nuclear Information System (INIS)

Yusuf-Nampira; Narko-Wibowo, L; Rosika-Krisnawati; Nudia-Barenzani

2000-01-01

The using of x-ray fluorescence in the chemical analysis depend heavilyon the parameters of x-ray chamber, for examples : electric voltage andelectric current. That parameter give effect in the result of determine ofSn, Cr, Fe and Ni in the zirconium alloy. 20 kV electric voltages are used onthe Mo x-ray chamber shall product x-ray of zirconium in the sample materialcan give effect in the stability of the analysis result (deviation more than5%). The result of analysis of elements in the zirconium alloy shall givedeviation less than 5% when using of electric voltage of the x-ray chamberless than 19 kV. The sensitivity of analysis can be reached by step upelectric current of x-ray chamber. (author)

Pitting morphologies of zirconium base alloys in aqueous and non aqueous chloride media

International Nuclear Information System (INIS)

Palit, G.C.; Gadiyar, H.S.

1988-01-01

Pitting morphology of zirconium and Zr-Cr alloys in aqueous chloride and nonaqueous methanol + 0.4 per cent HCl solution was investigated and observed to follow different modes in these two environments. While in aqueous chloride solution pitting was transgranular and randomly oriented, in methanol-chloride solution pits were observed to initiate and propagate along the grain boundaries. In aqueous chloride solution very irregular and sponge like zirconium metal was formed inside the pit while in methanol-chloride solution the pits were crystallographic in nature. Optical microscopy has revealed that pits preferentially initiate and propagate along scratch line in aqueous chloride solution, but such was not the case in nonaqueous methanol-chloride solution. The nature and the mechanism operating in the catastropic failure of these materials are investigated. (author). 10 refs., 11 figs

MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

Directory of Open Access Journals (Sweden)

Miroslav Cech

2016-12-01

Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

Final report on: Grain size determination in zirconium alloys (IAEA Research Contract No. 6025/Rb.)

International Nuclear Information System (INIS)

Martinez M, E.

1991-12-01

In spite of the amount of research developed the knowledge still is far from complete and in this basis the International Atomic Energy Agency, (IAEA), by means of the Working Group on Water Reactor Fuel Performance and Technology, initiated, in 1990 the Coordinated Research Programme named Grain Size Determination In Zirconium Alloys. Several countries were invited to participate and to contribute to the main objective of the programme, which can be state as: To develop a unified metallographic technique capable to show the microstructure of zircaloy in a reproducible and uniform manner. To fulfill the objective the following goals were established: A. To measure the grain size and perform an statistical treatment, in samples prepared specifically to show different amounts of cold work, recrystallization and grain growth. B. To compare the results obtained by the different laboratories involved in the programme. C. Finally, after the Ugine meeting, also the determination of the recrystallization and grain growth kinetics. (Author)

Hydride formation on deformation twin in zirconium alloy

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju-Seong [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Sung-Dae [Korea Institute of Material Science (KIMS), 797 Changwondaero, Changwon, Gyeongnam, 642-831 (Korea, Republic of); Yoon, Jonghun, E-mail: yooncsmd@gmail.com [Department of Mechanical Engineering, Hanyang University, 1271 Sa3-dong, Sangrok-gu, Ansan-si, Gyeonggi-do, 426-791 (Korea, Republic of)

2016-12-15

Hydrides deteriorate the mechanical properties of zirconium (Zr) alloys used in nuclear reactors. Intergranular hydrides that form along grain boundaries have been extensively studied due to their detrimental effects on cracking. However, it has been little concerns on formation of Zr hydrides correlated with deformation twins which is distinctive heterogeneous nucleation site in hexagonal close-packed metals. In this paper, the heterogeneous precipitation of Zr hydrides at the twin boundaries was visualized using transmission electron microscopy. It demonstrates that intragranular hydrides in the twinned region precipitates on the rotated habit plane by the twinning and intergranular hydrides precipitate along the coherent low energy twin boundaries independent of the conventional habit planes. Interestingly, dislocations around the twin boundaries play a substantial role in the nucleation of Zr hydrides by reducing the misfit strain energy.

The effect of substrate texture and oxidation temperature on oxide texture development in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Garner, A., E-mail: alistair.garner@manchester.ac.uk [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom); Frankel, P. [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom); Partezana, J. [Westinghouse Electric Company, 1332 Beulah Road, Pittsburgh, PA 15235 (United States); Preuss, M. [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom)

2017-02-15

During corrosion of zirconium alloys a highly textured oxide is formed, the degree of this preferred orientation has previously been shown to be an important factor in determining the corrosion behaviour of these alloys. Two distinct experiments were designed in order to investigate the origin of this oxide texture development on two commercial alloys. Firstly, sheet samples of Zircaloy-4 were oxidised between 500 and 800 °C in air. The resulting monoclinic oxide texture strength was observed to decrease with increasing oxidation temperature. In a second experiment, orthogonal faces of Low Tin ZIRLO{sub ™} were oxidised in 360 °C water, providing different substrate textures but identical microstructures. The substrate texture was observed to have a negligible effect on the corrosion performance whilst the major orientation of both oxide phases was found to be independent of substrate orientation. It is concluded that the main driving force for oxide texture development in single-phase zirconium alloys is the compressive stress caused by the Zr−ZrO{sub 2} transformation. - Highlights: • Substrate orientation does not significantly affect oxide texture development. • Corrosion performance is independent of substrate texture. • Monoclinic oxide texture strength decreases with increasing oxidation temperature. • The main driving force for texture development is the oxidation-induced stress.

Process and equipement for zone heat treatment of zirconium alloys tubes

International Nuclear Information System (INIS)

Kiesler, A.J.; Frischmann, P.G.; Rockwood, A.C.

1977-01-01

Process for the thermal treatment of an area of a long zirconium alloy part in order to increase its corrosion resistance in the cooling conditions of boiling water reactor, in which the part is moved lengthwise through a succession of critical maximum temperature areas, during a critical time and is subjected to a temperature reduction at critical rate, so that each successive portion reaches a maximum temperature between 825 0 C, and directing water at a temperature around 60 to 80 0 C as jets in the cooling area [fr

Microstructural modelling and lubrication study during zirconium alloy hot extrusion

International Nuclear Information System (INIS)

Gaudout, B.

2009-01-01

Using torsion tests (with strain rate jumps) and an experimental hot mini-extrusion apparatus, several samples zirconium alloy have been deformed: Zircaloy-4 (high α range) and Zr-1Nb (α + β domain). The fragmentation of the microstructure and post-dynamic grain growth have been examined. The main difference between these two alloys is that Zr-1Nb does not show grain growth during a heat treatment within the α + β domain after hot deformation. The recrystallization volume fraction has been measured on extruded samples with or without heat treatment. These rheological and microstructural data have been used to determine the parameters of a microstructural model including: a work-hardening model (Laaasraoui/Jonas), a continuous dynamic recrystallization model (Gourdet/Montheillet) and a grain growth model. This model leads to a good prediction of recrystallization volume fraction for Zircaloy-4 extrusion. However, the Zr-1Nb model cannot be validated because of the difficulty to observe deformed microstructures. Extrusion process is lubricated with a solid film. Trapping tests show that this lubricant is thermoviscoplastic. Friction along the container and several observations show the lubrication is not realized by a continuous film. Indeed, the heterogeneousness of deformation of these alloys causes a rupture of the lubricant film. Experiments and numerical simulations show that the radial gradient of axial displacement is affected by friction but also by stress softening of the alloys. (author)

Experimental specifications for eutectic reaction between metallic fuel and HT-9

International Nuclear Information System (INIS)

Hwang, Woan; Nam, Cheol; Lee, Byoung Oon; Ryu, Woo Seog

1998-10-01

The chemical interaction between metallic fuel and cladding is important in designing the fuel pin of the KALIMER. When metal fuel and cladding are contacted, the elements in fuel and cladding are inter-diffuse each other, forming the reaction layers at interface. The reaction layers may cause two important factors in aspects of fuel pin integrity. Firstly, it degrades cladding strength by reducing effective cladding thickness. Secondly, these layers accelerate eutectic reaction at transient conditions. To evaluate these phenomena, the diffusion couple experiment is planned by using metal fuels with various zirconium contents and HT-9 steel. The U-Zr fuel alloys will be used for the experiment with the different zirconium contents, these are 8, 10 and 12 weight %. This experiment aims to evaluate the effects of zirconium content on the chemical reaction. Furthermore, the reaction rate and threshold temperature of the eutectic melting will be determined as a function of the zirconium content. This document describes the detail experimental specifications for the eutectic reaction such as test setup, test requirements and test procedure. (author). 10 refs

Development of Silicide Coating on Molybdenum Alloy Cladding

International Nuclear Information System (INIS)

Lim, Woojin; Ryu, Ho Jin

2015-01-01

The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

High-temperature thermodynamic activities of zirconium in platinum alloys determined by nitrogen-nitride equilibria

International Nuclear Information System (INIS)

Goodman, D.A.

1980-05-01

A high-temperature nitrogen-nitride equilibrium apparatus is constructed for the study of alloy thermodynamics to 2300 0 C. Zirconium-platinum alloys are studied by means of the reaction 9ZrN + 11Pt → Zr 9 Pt 11 + 9/2 N 2 . Carful attention is paid to the problems of diffusion-limited reaction and ternary phase formation. The results of this study are and a/sub Zr//sup 1985 0 C/ = 2.4 x 10 -4 in Zr 9 Pt 11 ΔG/sub f 1985 0 C/ 0 Zr 9 Pt 11 less than or equal to -16.6 kcal/g atom. These results are in full accord with the valence bond theory developed by Engel and Brewer; this confirms their prediction of an unusual interaction of these alloys

Atomistic studies of cation transport in tetragonal ZrO2 during zirconium corrosion

International Nuclear Information System (INIS)

Bai, Xian-Ming; Zhang, Yongfeng; Tonks, Michael R.

2015-01-01

Zirconium alloys are the major fuel cladding materials in current reactors. The water-side corrosion is a significant degradation mechanism of these alloys. During corrosion, the transport of oxidizing species in zirconium dioxide (ZrO 2 ) determines the corrosion kinetics. Previously, it has been argued that the outward diffusion of cations is important for forming protective oxides. In this work, the migration of Zr defects in tetragonal ZrO 2 is studied with temperature accelerated dynamics and molecular dynamics simulations. The results show that Zr interstitials have anisotropic diffusion and migrate preferentially along the [001] or c direction in tetragonal ZrO 2 . The compressive stresses can increase the Zr interstitial migration barrier significantly. The migration of Zr interstitials at a grain boundary is much slower than in a bulk oxide. The implications of these atomistic simulation results in the Zr corrosion are discussed. (authors)

Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Steele, W.G.; Wazzan, A.R.; Okrent, D.

1989-01-01

An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Ribis, J.

2007-12-01

The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

Modelling of stress corrosion cracking in zirconium alloys

International Nuclear Information System (INIS)

Fandeur, O.; Rouillon, L.; Pilvin, P.; Jacques, P.; Rebeyrolle, V.

2001-01-01

During normal and incidental operating conditions, PWR power plants must comply with the first safety requirement, which is to ensure that the cladding wall is sound. Indeed some severe power transients potentially induce Stress Corrosion Cracking (SCC) of the zirconium alloy clad, due to strong Pellet Cladding Interaction (PCI). Since, at present, the prevention of this risk has some consequences on the French reactors manoeuvrability, a better understanding and forecast of the clad damage related to SCC/PCI is needed. With this aim, power ramp tests are performed in experimental reactors to assess the fuel rod behaviour and evaluate PCI failure risks. To study in detail SCC mechanisms, additional laboratory experiments are carried out on non-irradiated and irradiated cladding tubes. Numerical simulations of these tests have been developed aiming, on the one hand, to evaluate mechanical state variables and, on the other hand, to study consistent mechanical parameters for describing stress corrosion clad failure. The main result of this simulation is the determination of the validity ranges of the stress intensity factor, which is frequently used to model SCC. This parameter appears to be valid only at the onset of crack growth, when crack length remains short. In addition, the role of plastic strain rate and plastic strain as controlling parameters of the SCC process has been analysed in detail using the above mechanical description of the crack tip mechanical fields. Finally, the numerical determination of the first-order parameter(s) in the crack propagation rate law is completed by the development of laboratory tests focused on these parameters. These tests aim to support experimentally the results of the FE simulation. (author)

Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

International Nuclear Information System (INIS)

Harmon, H.D.

1975-01-01

A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

Oxidation behaviour of zirconium alloys and their precipitates – A mechanistic study

International Nuclear Information System (INIS)

Proff, C.; Abolhassani, S.; Lemaignan, C.

2013-01-01

The precipitate oxidation behaviour of binary zirconium alloys containing 1 wt.% Fe, Ni, Cr or 0.6 wt.% Nb was characterised in TEM on FIB prepared transverse sections of the oxide and reported in previous studies [1,2]. In the present study the following alloys: Zr1%Cu, Zr0.5%Cu0.5%Mo and pure Zr are analysed to add to the available information. In all cases, the observed precipitate oxidation behaviour in the oxide close to the metal-oxide interface could be described either with delayed oxidation with respect to the matrix or simultaneous oxidation as the surrounding zirconium matrix. Attempt was made to explain these observations, with different parameters such as precipitate size and structure, composition and thermodynamic properties. It was concluded that the thermodynamics with the new approach presented could explain most precisely their behaviour, considering the precipitate stoichiometry and the free energy of oxidation of the constituting elements. The surface topography of the oxidised materials, as well as the microstructure of the oxide presenting microcracks have been examined. A systematic presence of microcracks above the precipitates exhibiting delayed oxidation has been found; the height of these crack calculated using the Pilling–Bedworth ratios of different phases present, can explain their origin. The protrusions at the surface in the case of materials containing large precipitates can be unambiguously correlated to the presence of these latter, and the height can be correlated to the Pilling–Bedworth ratios of the phases present as well as the diffusion of the alloying elements to the surface and their subsequent oxidation. This latter behaviour was much more considerable in the case of Fe and Cu with Fe showing systematically diffusion to the outer surface.

Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963); Recherche d'alliages de zirconium compatibles avec le gaz carbonique sous pression jusqu'a 500 ou 600 deg C (1063)

Energy Technology Data Exchange (ETDEWEB)

Baque, P; Dominget, R; Bossard, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO{sub 2}-cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [French] Le zirconium se trouve parmi les metaux a section de capture neutronique relativement faible et possede une temperature de fusion elevee; aussi peut on songer a l'employer notamment comme materiau de gainage d'elements combustibles pour reacteurs nucleaires refroidis au gaz carbonique. Une etude prealable de plusieurs qualites de zirconium a montre que le metal est deja assez fortement oxyde dans ce gaz des 500 deg C. En effet, le phenomene de ''breakaway'' est general; la vitesse d'oxydation devient alors lineaire et depend de la pression du gaz carbonique. La recherche d'alliages binaires et ternaires a donc ete entreprise afin de tenter d'ameliorer le comportement du metal. Elle a permis d'aboutir a quelques compositions interessantes: cuivre 1, 1,6 et 2,5 pour cent, vanadium 2 pour cent, et calcium 0,05 et 0,5 pour cent. Des alliages ternaires au cuivre-molybdene et cuivre-phosphore sont egalement moins oxydables, et en particulier ne presentent pas le phenomene de ''breakaway'', meme apres une longue exposition a 600 deg C. (auteurs)

Problems of zirconium metal production in Czechoslovakia

International Nuclear Information System (INIS)

Vareka, J.; Vaclavik, E.

1975-01-01

The problems are summed up of the production and quality control of zirconium sponge. A survey is given of industrial applications of zirconium in form of pure metal or alloys in nuclear power production, ferrous and non-ferrous metallurgy, chemical engineering and electrical engineering. A survey is also presented of the manufacture of zirconium metal in advanced capitalist countries. (J.B.)

Evaluation of Conditions for Hydrogen Induced Degradation of Zirconium Alloys during Fuel Operation and Storage. Final Report of a Coordinated Research Project 2011-2015

International Nuclear Information System (INIS)

2015-12-01

This publication reports on the work carried out in 2011–2015 in the coordinated research project (CRP) on the evaluation of conditions for hydrogen induced degradation of zirconium alloys during fuel operation and storage. The CRP was carried out to evaluate the threshold condition for delayed hydride cracking (KIH) in pressurized water reactors and zircaloy-4 and E635M fuel claddings, with application to in-pile operation and spent fuel storage. The project consisted of adding hydrogen to samples of cladding and measuring K IH by one of four methods. The CRP was the third in the series, of which the results of the first two were published in IAEA-TECDOC-1410 and IAEA-TECDOC-1649, in 2004 and 2010, respectively. This publication includes all of the research work performed in the framework of the CRP, including details of the experimental procedures that led to a set of data for tested materials. The research was conducted by representatives from 13 laboratories from all over the world. In addition to the basic goal to transfer the technology of the testing techniques from experienced laboratories to those unfamiliar with the methods, the CRP was set up to develop experimental procedures to produce consistent sets of data, both within a single laboratory and among different laboratories. The material condition and temperature history were prescribed, and laboratories chose one or two of four methods of loading that were recommended in an attempt to develop standard sets of experimental protocols so that consistent results could be obtained. Experimental discrepancies were minimized through careful attention to details of microstructure, temperature history and stress state in the samples, with the main variation being the mode of loading

Processing fissile material mixtures containing zirconium and/or carbon

Science.gov (United States)

Johnson, Michael Ernest; Maloney, Martin David

2013-07-02

A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

Determination of hydrogen in uranium-niobium-zirconium alloy by inert-gas fusion

International Nuclear Information System (INIS)

Carden, W.F.

1979-12-01

An improved method has been developed using inert-gas fusion for determining the hydrogen content in uranium-niobium-zirconium (U-7.5Nb-2.5Zr) alloy. The method is applicable to concentrations of hydrogen ranging from 1 to 250 micrograms per gram and may be adjusted for analysis of greater hydrogen concentrations. Hydrogen is determined using a hydrogen determinator. The limit of error for a single determination at the 95%-confidence level (at the 3.7-μg/g-hydrogen level) is +-1.4 micrograms per gram hydrogen

X-ray study of texture in zirconium alloy tubes and in graphite

International Nuclear Information System (INIS)

Skvortsov, V.V.; Alekseev, S.I.

1987-01-01

X-ray study of texture in zirconium alloy tubes and in graphite has been developed. The method is based on constructing coordinate grid of stereographic projection determining quantity and coordinates of points where measurements should be performed depending on a specimen slope pitch. Complete stereographic projection obtained so is a base both for constructing pole figures showing distribution normales of plane system being studied and for calculating texture coefficients determining property anisotropy in materials under investigation. This method can be applied to study texture in items of any materials independent of the item shape

Effect of zirconium addition on the recrystallization behaviour of a ...

Indian Academy of Sciences (India)

In the present work, zirconium was added to a commercial Al–Cu–Mg alloy and by heat treatment Al3Zr particles were precipitated and after forging, the grain size was an order of magnitude lower than the alloy without zirconium. Transmission electron microscopy was employed to characterize the second phase particles, ...

Numerical assessment of bone remodeling around conventionally and early loaded titanium and titanium-zirconium alloy dental implants.

Science.gov (United States)

Akça, Kıvanç; Eser, Atılım; Çavuşoğlu, Yeliz; Sağırkaya, Elçin; Çehreli, Murat Cavit

2015-05-01

The aim of this study was to investigate conventionally and early loaded titanium and titanium-zirconium alloy implants by three-dimensional finite element stress analysis. Three-dimensional model of a dental implant was created and a thread area was established as a region of interest in trabecular bone to study a localized part of the global model with a refined mesh. The peri-implant tissues around conventionally loaded (model 1) and early loaded (model 2) implants were implemented and were used to explore principal stresses, displacement values, and equivalent strains in the peri-implant region of titanium and titanium-zirconium implants under static load of 300 N with or without 30° inclination applied on top of the abutment surface. Under axial loading, principal stresses in both models were comparable for both implants and models. Under oblique loading, principal stresses around titanium-zirconium implants were slightly higher in both models. Comparable stress magnitudes were observed in both models. The displacement values and equivalent strain amplitudes around both implants and models were similar. Peri-implant bone around titanium and titanium-zirconium implants experiences similar stress magnitudes coupled with intraosseous implant displacement values under conventional loading and early loading simulations. Titanium-zirconium implants have biomechanical outcome comparable to conventional titanium implants under conventional loading and early loading.

A Novel Zr-1Nb Alloy and a New Look at Hydriding

Energy Technology Data Exchange (ETDEWEB)

Robert D. Mariani; James I. Cole; Assel Aitkaliyeva

2013-09-01

A novel Zr-1Nb has begun development based on a working model that takes into account the hydrogen permeabilities for zirconium and niobium metals. The beta-Nb secondary phase particles (SPPs) in Zr-1Nb are believed to promote more rapid hydrogen dynamics in the alloy in comparison to other zirconium alloys. Furthermore, some hydrogen release is expected at the lower temperatures corresponding to outages when the partial pressure of H2 in the coolant is less. These characteristics lessen the negative synergism between corrosion and hydriding that is otherwise observed in cladding alloys without niobium. In accord with the working model, development of nanoscale precursors was initiated to enhance the performance of existing Zr-1Nb alloys. Their characteristics and properties can be compared to oxide-dispersion strengthened alloys, and material additions have been proposed to zirconium-based LWR cladding to guard further against hydriding and to fix the size of the SPPs for microstructure stability enhancements. A preparative route is being investigated that does not require mechanical alloying, and 10 nanometer molybdenum particles have been prepared which are part of the nanoscale precursors. If successful, the approach has implications for long term dry storage of used fuel and for new routes to nanoferritic and ODS alloys.

Characterization of IFR metal fuel fragmentation

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1987-01-01

The integral fast reactor (IFR) employs a reactor design that has inherent safety features. An important safety advantage is derived from its pool configuration, which facilitates passive decay heat removal and isolates the core from accidents that might occur elsewhere in the plant. The metal-alloy fuel has superior heat transfer properties compared to oxide fuels. While the IFR design has these inherent safety features, a complete analysis of reactor safety requires assessment of the consequences of the melting of the uranium alloy fuel in the core and the contact of molten core materials with sodium. A series of eight tests was conducted in which the fragmentation and interaction behavior of kilogram quantities of uranium-zirconium alloy in sodium was studied

Characterisation of hydrides in a zirconium alloy, by EBSD

International Nuclear Information System (INIS)

Ubhi, H.S.; Larsen, K.

2012-01-01

Zirconium alloys are used in nuclear reactors owing to their low capture cross-section for thermal neutrons and good mechanical and corrosion properties. However, they do suffer from delayed hydrogen cracking (DHC) due to formation of hydride particles. This study shows how the electron back-scatter diffraction (EBSD) technique can be used to characterise hydrides and their orientation relationship with the matrix. Hydrided EB weld specimens were prepared by electro-polishing, characterised using Oxford instruments AZtecHKL EBSD apparatus and software attached to a FEG SEM. Hydrides were found to exist as fine intra granular plates and having the Blackburn orientation relationship, i.e. (0002)Zr//(111)hydride and (1120)Zr//(1-10)hydride. The hydrides were also found to contain sigma 3 boundaries as well as local misorientations. (author)

Process for etching zirconium metallic objects

International Nuclear Information System (INIS)

Panson, A.J.

1988-01-01

In a process for etching of zirconium metallic articles formed from zirconium or a zirconium alloy, wherein the zirconium metallic article is contacted with an aqueous hydrofluoric acid-nitric acid etching bath having an initial ratio of hydrofluoric acid to nitric acid and an initial concentration of hydrofluoric and nitric acids, the improvement, is described comprising: after etching of zirconium metallic articles in the bath for a period of time such that the etching rate has diminished from an initial rate to a lesser rate, adding hydrofluoric acid and nitric acid to the exhausted bath to adjust the concentration and ratio of hydrofluoric acid to nitric acid therein to a value substantially that of the initial concentration and ratio and thereby regenerate the etching solution without removal of dissolved zirconium therefrom; and etching further zirconium metallic articles in the regenerated etching bath

Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

Energy Technology Data Exchange (ETDEWEB)

Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

Dependency of Delayed Hydride Crack Velocity on the Direction of an Approach to Test Temperatures in Zirconium Alloys

International Nuclear Information System (INIS)

Kim, Young Suk; Kim, Kang Soo; Im, Kyung Soo; Ahn, Sang Bok; Cheong, Yong Moo

2005-01-01

Recently, Kim proposed a new DHC model where a driving force for the DHC is a supersaturated hydrogen concentration as a result of a hysteresis of the terminal solid solubility (TSS) of hydrogen in zirconium alloys upon a heating and a cooling. This model was demonstrated to be valid through a model experiment where the prior plastic deformation facilitated nucleation of the reoriented hydrides, thus reducing the supersaturated hydrogen concentration at the plastic zone ahead of the crack tip and causing hydrogen to move to the crack tip from the bulk region. Thus, an approach to the test temperature by a cooling is required to create a supersaturation of hydrogen, which is a driving force for the DHC of zirconium alloys. However, despite the absence of the supersaturation of hydrogen due to an approach to the test temperature by a heating, DHC is observed to occur in zirconium alloys at the test temperatures below 180 .deg. C. As to this DHC phenomenon, Kim proposed that stress-induced transformation from γ-hydrides to δ-hydrides is likely to be a cause of this, based on Root's observation that the γ-hydride is a stable phase at temperatures lower than 180 .deg. C. In other words, the hydrides formed at the crack tip would be δ-hydrides due to the stressinduced transformation while the bulk region still maintains the initial hydride phase or γ-hydrides. It should be noted that Ambler has also assumed the crack tip hydrides to be δ-hydrides. When the δ-hydrides or ZrH1.66 are precipitated at the crack tip due to the transformation of the γ-hydrides or ZrH, the crack tip will have a decreased concentration of dissolved hydrogen in zirconium, considering the atomic ratio of hydrogen and zirconium in the γ- and δ-hydrides. In contrast, due to no stress-induced transformation of hydrides, the bulk region maintains the initial concentration of dissolved hydrogen. Hence, there develops a difference in the hydrogen concentration or .C between the bulk and the

Dependency of Delayed Hydride Crack Velocity on the Direction of an Approach to Test Temperatures in Zirconium Alloys

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kim, Kang Soo; Im, Kyung Soo; Ahn, Sang Bok; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

Recently, Kim proposed a new DHC model where a driving force for the DHC is a supersaturated hydrogen concentration as a result of a hysteresis of the terminal solid solubility (TSS) of hydrogen in zirconium alloys upon a heating and a cooling. This model was demonstrated to be valid through a model experiment where the prior plastic deformation facilitated nucleation of the reoriented hydrides, thus reducing the supersaturated hydrogen concentration at the plastic zone ahead of the crack tip and causing hydrogen to move to the crack tip from the bulk region. Thus, an approach to the test temperature by a cooling is required to create a supersaturation of hydrogen, which is a driving force for the DHC of zirconium alloys. However, despite the absence of the supersaturation of hydrogen due to an approach to the test temperature by a heating, DHC is observed to occur in zirconium alloys at the test temperatures below 180 .deg. C. As to this DHC phenomenon, Kim proposed that stress-induced transformation from {gamma}-hydrides to {delta}-hydrides is likely to be a cause of this, based on Root's observation that the {gamma}-hydride is a stable phase at temperatures lower than 180 .deg. C. In other words, the hydrides formed at the crack tip would be {delta}-hydrides due to the stressinduced transformation while the bulk region still maintains the initial hydride phase or {gamma}-hydrides. It should be noted that Ambler has also assumed the crack tip hydrides to be {delta}-hydrides. When the {delta}-hydrides or ZrH1.66 are precipitated at the crack tip due to the transformation of the {gamma}-hydrides or ZrH, the crack tip will have a decreased concentration of dissolved hydrogen in zirconium, considering the atomic ratio of hydrogen and zirconium in the {gamma}- and {delta}-hydrides. In contrast, due to no stress-induced transformation of hydrides, the bulk region maintains the initial concentration of dissolved hydrogen. Hence, there develops a difference in the

The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

International Nuclear Information System (INIS)

Sawatzky, A.

1964-10-01

Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

The quest for safe and reliable fuel cladding materials

Energy Technology Data Exchange (ETDEWEB)

Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

The quest for safe and reliable fuel cladding materials

International Nuclear Information System (INIS)

Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia

2015-01-01

The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

Nuclear fuel element and a method of manufacture thereof

International Nuclear Information System (INIS)

Wood, J.C.

1975-01-01

A nuclear fuel element having a sheath of zirconium or a zirconium alloy and a cross-linked siloxane lacquer coating on the inner surface of the sheath and separating the nuclear fuel material from the sheath is described. The siloxane lacquer coating retards cracking of the sheath by iodine vapor emitted by the fuel during burn-up, and acts as a lubricant for the fuel to prevent rupture of the sheath by thermal ratchetting of the fuel against the sheath and caused by differential thermal expansion between the fuel and the sheath. A silicone grease is applied as a thin layer in the sheath and then baked so that oxidative cleavage of the side chains of the grease occurs to form a cross-linked siloxane lacquer coating bonded to the sheath

Nuclear fuel element

International Nuclear Information System (INIS)

Iwano, Yoshihiko.

1993-01-01

Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

Quantitative analysis of nickel in zirconium and zircaloy

International Nuclear Information System (INIS)

Rastoix, M.

1957-01-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [fr

Corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

1993-01-01

To improve our understanding of corrosion mechanisms under irradiation of zirconium alloys, to collect information systematically and to identify areas where further experimentation is needed, in 1989 the IAEA initiated a special project with the participation of expert from Canada, France, Japan, USA and the former USSR. This technical document is the result of two years of joint investigations. In view of the rapidly evolving mechanistic understanding of the phenomena in this field, the document presents a series of snapshots of current ideas in specific areas of study that are relevant to the whole problem. Any attempt to present an agreed upon micromechanistic hypothesis that explains the overall phenomena must await further detailed investigations. Throughout the text, the authors have endeavored to indicate critical gaps in our basic knowledge. It is hoped that this will stimulate experimental studies in just those areas where further data are most urgently required. Refs, figs and tabs

Corrosion behavior of Zr-x(Nb, Sn and Cu) binary alloys

International Nuclear Information System (INIS)

Kim, M. H.; Lee, M. H.; Park, S. Y.; Jung, Y. H.; We, M. Y.

1999-01-01

For the development of advanced zirconium alloys for nuclear fuel cladding, the corrosion behaviors of zirconium binary alloys were studied on the Zr-xNb, Zr-xSn, and Zr-xCu alloys. The corrosion test were performed in water at 360 deg C, steam at 400 deg C and LiOH at 360 deg C for 45 days. The corrosion behaviors of Zr-xNb was similar to that of Zr-xCu alloys. However, the corrosion behavior of Zr-xSn was different from Zr-xNb and Zr-xCu. The weight gain of Zr-xNb and Zr-xCu was increased with addition of alloying elements. When Sn is added to Zr matrix in range below the solubility limit, the corrosion resistance decrease with increasing Sn-content, while in the range over solubility limit, Sn has an adverse effect on the corrosion resistance. Especially, Zr-xSn alloys showed higher corrosion resistance than Zr-xNb and Zr-xCu alloys in LiOH solution

Lithium uptake and the accelerated corrosion of zirconium alloys

International Nuclear Information System (INIS)

Ramasubramanian, N.; Precoanin, N.; Ling, V.C.

1989-01-01

The corrosion of zirconium alloys in aqueous lithiated solutions is sensitive to the concentration of the alkali and the temperature. In concentrated solutions, >10 -1 M in lithium hydroxide (LiOH) (700-ppm lithium) and at temperatures >573 K, accelerated corrosion occurs at quite an early stage. Our investigations indicate that the accelerated corrosion is caused by the generation of porosity, rather than the dissolution of lithium, in the growing oxide. Specimens of standard Zircaloy-4 fuel cladding and Zr-2.5 wt% Nb pressure tube materials were corroded in lithium hydroxide solutions, 10 -3 to 1 M in concentration, at 589 K. Impedance measurements, polarizations in molten lithium nitrate-lithium hydroxide (LiNO 3 -LiOH) and scanning electron microscopy of the alloy-oxide interface indicated a high level of porosity, right from the initial stages, for oxide films grown in the concentrated solutions. The oxides, when analyzed by atomic absorption spectroscopy, revealed the presence of a few 100 ppm of lithium, too small to account for the accelerated corrosion by a mechanism of solid solution of lithium in zirconia. X-ray powder patterns of the oxides showed peaks for only monoclinic zirconia, but occasionally peaks for LiOH · H 2 O and LiOH were also observed. The counts for lithium, detected by secondary ion mass spectrometry, decreased when specimens cut from the same corroded samples were leached in nitric acid. It is concluded from these observations that a major part of lithium is physically held in the porous oxide. Lithium hydroxide is not completely dissociated in aqueous solutions; with increasing concentration and temperature, an increasingly larger proportion of the alkali remains undissociated. It is suggested that the accelerated corrosion in concentrated solutions is caused by the participation of the undissociated alkali in the reactions occurring on the surfaces of the zirconia crystallites. The undissociated LiOH and hydroxyl ions react at an

An X-ray absorption near-edge structure (XANES) study of the Sn L_3 edge in zirconium alloy oxide films formed during autoclave corrosion

International Nuclear Information System (INIS)

Hulme, Helen; Baxter, Felicity; Babu, R. Prasath; Denecke, Melissa A.; Gass, Mhairi; Steuwer, Axel; Norén, Katarina; Carlson, Stefan; Preuss, Michael

2016-01-01

Highlights: • Characterisation of tin speciation in zirconium alloy metal and oxide films using Sn L_3-XANES. • Chemical environment of tin in Zircaloy-4 and ZIRLO™ oxide films shown to be similar. • Tin in the oxide films is present in both the di- and tetravalent states and oxidises progressively with oxide-layer growth. - Abstract: Application of Sn L_3-XANES to study the oxidation state of alloying additions of tin (1–1.2 wt%) in <2 μm oxide layers formed on nuclear grade zirconium alloy has been demonstrated. Data obtained for metallic and corroded ZIRLO™ (1 wt% Sn) and Zircaloy-4 (1.2 wt% Sn) indicate tin has a similar chemical speciation in both metal alloys but this differs in the oxidised surface layers. By recording XANES at various incident angles to vary the photon penetration depth and amount of the oxide layer probed in the measurement, the authors found evidence that the oxidation of tin progresses with increasing oxide thickness.

Accident tolerant fuel cladding development: Promise, status, and challenges

Science.gov (United States)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

Thin polycrystalline diamond films protecting zirconium alloys surfaces: From technology to layer analysis and application in nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Ashcheulov, P. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Škoda, R.; Škarohlíd, J. [Czech Technical University in Prague, Faculty of Mechanical Engineering, Technická 4, Prague 6, CZ-160 07 (Czech Republic); Taylor, A.; Fekete, L.; Fendrych, F. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Vega, R.; Shao, L. [Texas A& M University, Department of Nuclear Engineering TAMU-3133, College Station, TX TX 77843 (United States); Kalvoda, L.; Vratislav, S. [Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic); Cháb, V.; Horáková, K.; Kůsová, K.; Klimša, L.; Kopeček, J. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Sajdl, P.; Macák, J. [University of Chemistry and Technology, Power Engineering Department, Technická 3, Prague 6, CZ-166 28 (Czech Republic); Johnson, S. [Nuclear Fuel Division, Westinghouse Electric Company, 5801 Bluff Road, Hopkins, SC 29209 (United States); Kratochvílová, I., E-mail: krat@fzu.cz [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic)

2015-12-30

Graphical abstract: - Highlights: • In this work we showed that films prepared by MW-LA-PECVD technology can be used as anticorrosion protective layer for Zircaloy2 nuclear fuel claddings at elevated temperatures (950 °C) when α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). Quality of PCD films was examined by Raman spectroscopy, XPS, SEM, AFM and SIMS analysis. • The polycrystalline diamond films were of high quality - without defects and contaminations. After hot steam oxidation (950 °C) a high level of structural integrity of PCD layer was observed. Both sp{sup 2} and sp{sup 3} C phases were present in the protective PCD layer. Higher resistance and a lower degree of impedance dispersion was found in the hot steam oxidized PCD coated Zircaloy2 samples, which may suggest better protection of the Zircaloy2 surface. The PCD layer blocks the hydrogen diffusion into the Zircaloy2 surface thus protecting the material from degradation. • Hot steam oxidation tests confirmed that PCD coated Zircaloy2 surfaces were effectively protected against corrosion. Presented results demonstrate that the PCD anticorrosion protection can significantly prolong service life of Zircaloy2 nuclear fuel claddings in nuclear reactors even at elevated temperatures. - Abstract: Zirconium alloys can be effectively protected against corrosion by polycrystalline diamond (PCD) layers grown in microwave plasma enhanced linear antenna chemical vapor deposition apparatus. Standard and hot steam oxidized PCD layers grown on Zircaloy2 surfaces were examined and the specific impact of polycrystalline Zr substrate surface on PCD layer properties was investigated. It was found that the presence of the PCD coating blocks hydrogen diffusion into the Zircaloy2 surface and protects Zircaloy2 material from degradation. PCD anticorrosion protection of Zircaloy2 can significantly prolong life of Zircaloy2 material in nuclear reactors even at temperatures above Zr

The effect of copper, chromium, and zirconium on the microstructure and mechanical properties of Al-Zn-Mg-Cu alloys

Science.gov (United States)

Wagner, John A.; Shenoy, R. N.

1991-01-01

The present study evaluates the effect of the systematic variation of copper, chromium, and zirconium contents on the microstructure and mechanical properties of a 7000-type aluminum alloy. Fracture toughness and tensile properties are evaluated for each alloy in both the peak aging, T8, and the overaging, T73, conditions. Results show that dimpled rupture essentially characterize the fracture process in these alloys. In the T8 condition, a significant loss of toughness is observed for alloys containing 2.5 pct Cu due to the increase in the quantity of Al-Cu-Mg-rich S-phase particles. An examination of T8 alloys at constant Cu levels shows that Zr-bearing alloys exhibit higher strength and toughness than the Cr-bearing alloys. In the T73 condition, Cr-bearing alloys are inherently tougher than Zr-bearing alloys. A void nucleation and growth mechanism accounts for the loss of toughness in these alloys with increasing copper content.

A new model for prediction of dispersoid precipitation in aluminium alloys containing zirconium and scandium

International Nuclear Information System (INIS)

Robson, J.D.

2004-01-01

A model has been developed to predict precipitation of ternary Al 3 (Sc, Zr) dispersoids in aluminium alloys containing zirconium and scandium. The model is based on the classical numerical method of Kampmann and Wagner, extended to predict precipitation of a ternary phase. The model has been applied to the precipitation of dispersoids in scandium containing AA7050. The dispersoid precipitation kinetics and number density are predicted to be sensitive to the scandium concentration, whilst the dispersoid radius is not. The dispersoids are predicted to enrich in zirconium during precipitation. Coarsening has been investigated in detail and it has been predicted that a steady-state size distribution is only reached once coarsening is well advanced. The addition of scandium is predicted to eliminate the dispersoid free zones observed in scandium free 7050, greatly increasing recrystallization resistance

Polarographic determination of the titanium and niobium content of zirconium alloys

International Nuclear Information System (INIS)

Levin, R; Gabra, J.

1978-03-01

A method is described for the polarographic determination of titanium and niobium in zirconium alloys in the concentration range of 0.1% to 4% of each of the determined metals. To assure the complete dissolution of the sample a mixture of nitric acid and hydrofluoric acid is used. After evaporating these acids in the presence of sulphuric acid, the contents are determined polarographically with a supporting electrolyte solution of 0.1M EDTA, 0.33M potassium sulfate and 0.4M sodium acetate, buffered to pH 4 with acetic acid. The half-wave potential (Esub(1/2)) of titanium is -0.35V and that of niobium is -0.67 V. (author)

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

Energy Technology Data Exchange (ETDEWEB)

Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-28

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

Metal waste forms from treatment of EBR-II spent fuel

International Nuclear Information System (INIS)

Abraham, D. P.

1998-01-01

Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

Chemical compatibility between cladding alloys and advanced fuels

International Nuclear Information System (INIS)

Fee, D.C.; Johnson, C.E.

1975-05-01

The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

Nuclear fuel element

International Nuclear Information System (INIS)

Armijo, J.S.

1977-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed which has a composite cladding having a substrate, a metal barrier metallurgically bonded to the inside surface of the substrate and an inner layer metallurgically bonded to the inside surface of the metal barrier. In this composite cladding, the inner layer and the metal barrier shield the substrate from any impurities or fission products from the nuclear fuel material held within the composite cladding. The metal barrier forms about 1 to about 4 percent of the thickness of the cladding and is comprised of a metal selected from the group consisting of niobium, aluminum, copper, nickel, stainless steel, and iron. The inner layer and then the metal barrier serve as reaction sites for volatile impurities and fission products and protect the substrate from contact and reaction with such impurities and fission products. The substrate and the inner layer of the composite cladding are selected from conventional cladding materials and preferably are a zirconium alloy. Also in a preferred embodiment the substrate and the inner layer are comprised of the same material, preferably a zirconium alloy. 19 claims, 2 figures

Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

Directory of Open Access Journals (Sweden)

Giovedi Claudia

2016-01-01

Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

On the initial corrosion mechanism of zirconium alloy: Interaction of oxygen and water with Zircaloy at room temperature and 450 C evaluated by x-ray absorption spectroscopy and photoelectron spectroscopy

International Nuclear Information System (INIS)

Doebler, U.; Knop, A.

1994-01-01

The initial stages of zirconium oxide formation on Zircaloy after water (H 2 O) and oxygen (O 2 ) exposures have been investigated in situ using photoelectron spectroscopy and X-ray-absorption spectroscopy. The reactivity of the zirconium alloy with O 2 at room temperature is about 1,000 times higher than for H 2 O. Up to 100 L (1 L = 1 Langmuir unit = 1 · 10 -6 mbar · s) H 2 O exposure, the reactivity of the zirconium alloy at 450 C is comparable to the room temperature reaction. At higher H 2 O exposure, a sharp increase in the reaction rate for the high-temperature oxidation is observed. From the energy position of the Zr 3d photo emission line and their oxygen-induced chemical shifts, one can really follow the formation of the oxide films. Two different substoichiometric oxides were found during reaction with water. Suboxide (1) is located at the zirconium/zirconium-oxide interface. Subsequently, a Suboxide (2) is concluded from the chemical shift of the zirconium photoelectrons. After an oxide thickness of 2 nm, the stoichiometric ZrO 2 phase is not yet developed

Influence of chemical composition of zirconium alloy E110 on embrittlement under LOCA conditions - Part 1: Oxidation kinetics and macrocharacteristics of structure and fracture

Science.gov (United States)

Nikulin, S. A.; Rozhnov, A. B.; Belov, V. A.; Li, E. V.; Glazkina, V. S.

2011-11-01

Exploratory investigations of the influence of alloying and impurity content in the E110 alloy cladding tubes on the behavior under conditions of Loss of Coolant Accidents (LOCA) has been performed. Three alloys of E110 type have been tested: E110 alloy of nominal composition Zr-1%Nb (E110), E110 alloy of modified composition Zr-1%Nb-0.12%Fe-0.13%O (E110M), E110 alloy of nominal composition Zr-1%Nb with reduced impurity content (E110G). Alloys E110 and E110M were manufactured on the electrolytic basis and alloy E110G was manufactured on the basis of zirconium sponge. The high temperature oxidation tests in steam ( T = 1100 °C, 18% of equivalent cladding reacted (ECR)) have been conducted, kinetics of oxidation was investigated. Quantitative research of structure and fracture macrocharacteristics was performed by means of optical and electron microscopy. The results received were compared with the residual ductility of specimens. The results of the investigation showed the existence of "breakaway oxidation" kinetics and white spalling oxide in E110 and E110M alloys while the specimen oxidation kinetics in E110G alloy was characterized by a parabolic law and specimens had a dense black oxide. Oxygen and iron alloying in the E110 alloy positively changed the macrocharacteristics of structure and fracture. However, in general, it did not improve the resistance to embrittlement in LOCA conditions apparently because of a strong impurity influence caused by electrolytic process of zirconium production.

Oxidation of zirconium-aluminum alloys

International Nuclear Information System (INIS)

Cox, B.

1967-10-01

Examination of the processes occurring during the oxidation of Zr-1% A1, Zr-3% A1, and Zr-1.5% A1-0.5% Mo alloys has shown that in steam rapid oxidation occurs predominantly around the Zr 3 A1 particles, which at low temperatures appear to be relatively unattacked. The unoxidised particles become incorporated in the oxide, and become fully oxidised as the film thickens. This rapid localised oxidation is preceded by a short period of uniform film growth, during which the oxide film thickness does not exceed ∼200A-o. Thus the high oxidation rates can probably be ascribed to aluminum in solution in the zirconium matrix, although its precise mode of operation has not been determined. Once the solubility limit of aluminum is exceeded, the size, distribution and number of intermetallic particles affects the oxidation rate merely by altering the distribution of regions of metal giving high oxidation rates. The controlling process during the early stages of oxidation is electron transport and not ionic transport. Thus, the aluminum in the oxide film is presumably increasing the ionic conductivity more than the electronic. The oxidation rates in atmospheric pressure steam are very high and their irregular temperature dependence suggests that the oxidation rate will be pressure dependent. This was confirmed, in part, by a comparison with oxidation in moist air. It was found that the rate of development of white oxide around intermetallic particles was considerably reduced by the decrease in the partial pressure of H 2 O; the incubation period was not much different, however. (author)

Fuel and fuel cycles with high burnup for WWER reactors

International Nuclear Information System (INIS)

Chernushev, V.; Sokolov, F.

2002-01-01

The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

Hardening of niobium alloys at precrystallization annealing

International Nuclear Information System (INIS)

Vasil'eva, E.V.; Pustovalov, V.A.

1989-01-01

Niobium base alloys were investigated. It is shown that precrystallization annealing of niobium-molybdenum, niobium-vanadium and niobium-zirconium alloys elevates much more sufficiently their resistance to microplastic strains, than to macroplastic strains. Hardening effect differs sufficiently for different alloys. The maximal hardening is observed for niobium-vanadium alloys, the minimal one - for niobium-zirconium alloys

Evaluation of delayed hydride cracking and fracture toughness in zirconium alloys

International Nuclear Information System (INIS)

Oh, Je Yong

2000-02-01

The tensile, fracture toughness, and delayed hydride cracking (DHC) test were carried at various temperatures to understand the effect of hydrides on zirconium alloys. And the effects of yield stress and texture on the DHC velocity were discussed. The tensile properties of alloy A were the highest, and the difference between directions in alloy C was small due to texture. The fracture toughness at room temperature decreased sharply when hydrided. Although the alignment of hydride plates was parallel to loading direction, the hydrides were fractured due to the triaxiality at the crack tip region. The fracture toughness over 200 .deg. C was similar regardless of the hydride existence, because the triaxiality region was lost due to the decrease of yield stress with temperature. As the yield stress decreased, the threshold stress intensity factor and the striation spacing increased in alloy A, and the fracture surfaces and striations were affected by microstructures in all alloys. To evaluate the effect of the yield stress on DHC velocity, a normalization method was proposed. When the DHC velocity was normalized with dividing by the terminal solid solubility and the diffusion coefficient of hydrogen, the relationship between the yield stress and the DHC velocity was representable on one master curve. The equation from the master curve was able to explain the difference between the theoretical activation energy and the experimental activation energy in DHC. The difference was found to be ascribed to the decrease of yield stress with temperature. texture affected the delayed hydride cracking velocity by yield stress and by hydride reprecipitation. The relationship between the yield stress and the DHC velocity was expressed as an exponential function, and the relationship between the reprecipitation of hydride and the DHC velocity was expressed as a linear function

Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

International Nuclear Information System (INIS)

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2007-01-01

For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

International Nuclear Information System (INIS)

Graff, St.

2006-10-01

The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

Energy Technology Data Exchange (ETDEWEB)

Graff, St

2006-10-15

The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling; Elaboration de zirconium par reduction de tetrachlorure de zirconium par magnesothermie. Etude experimentale et modelisation

Energy Technology Data Exchange (ETDEWEB)

Basin, N

2001-01-01

This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl{sub 4} + 2 Mg = 2 MgCl{sub 2}. By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

Hydrogen permeation in FeCrAl alloys for LWR cladding application

Science.gov (United States)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Study for the chlorination of zirconium oxide

International Nuclear Information System (INIS)

Seo, E.S.M.; Takiishi, H.; Paschoal, J.O.A.; Andreoli, M.

1990-12-01

In the development of new ceramic and metallic materials the chlorination process constitutes step in the formation of several intermediate compounds, such as metallic chlorides, used for the production of high, purity raw materials. Chlorination studies with the aim of fabrication special zirconium-base alloys have been carried out at IPEN. Within this program the chlorination technique has been used for zirconium tetrachloride production from zirconium oxide. In this paper some relevant parameters such as: time and temperature of reaction, flow rate of chloride gas and percentage of the reducing agent which influence the efficiency of chlorination of zirconium oxide are evaluated. Thermodynamical aspects about the reactions involved in the process are also presented. (author)

Fuel powder production from ductile uranium alloys

International Nuclear Information System (INIS)

Clark, C.R.; Meyer, M.K.

1998-01-01

Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Recovery of zirconium from pickling solution, regeneration and its reuse

Energy Technology Data Exchange (ETDEWEB)

Bhattacharjee, D. [Nuclear Fuel Complex, Hyderabad 500062 (India); Mandal, D., E-mail: dmandal10@gmail.com [Alkali Material & Metal Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Visweswara Rao, R.V.R.L.; Sairam, S.; Thakur, S. [Nuclear Fuel Complex, Hyderabad 500062 (India)

2017-05-15

Graphical abstract: The following compares the performance of fresh pickling solution (PS) and regenerated and used pickling solution (UPS). - Highlights: • Pickling of zircaloy tubes and appendages is carried out to remove oxide layer. • The pickling solution become saturated with zirconium due to reuse. • As NaNO{sub 3} concentration increases, conc. of Zr in pickling solution decreases. • Experimental results shows that, used pickling solution can be regenerated. • Regenerated solution may be reused by adding makeup quantities of HF-HNO{sub 3}. - Abstract: The pressurized heavy water reactors use natural uranium oxide (UO{sub 2}) as fuel and uses cladding material made up of zircaloy, an alloy of zirconium. Pickling of zircaloy tubes and appendages viz., spacer and bearing pads is carried out to remove the oxide layer and surface contaminants, if present. Pickling solution, after use for many cycles i.e., used pickling solution (UPS) is sold out to vendors, basically for its zirconium value. UPS, containing a relatively small concentration of hydrofluoric acid. After repeated use, pickling solution become saturated with zirconium fluoride complex and is treated by adding sodium nitrate to precipitate sodium hexafluro-zirconate. The remaining solution can be recycled after suitable makeup for further pickling use. The revenue lost by selling UPS is very high compared to its zirconium value, which causes monetary loss to the processing unit. Experiments were conducted to regenerate and reuse UPS which will save a good amount of revenue and also protect the environment. Experimental details and results are discussed in this paper.

Study on manufacturing technology of fuel guide tube using HANA alloys

International Nuclear Information System (INIS)

Kim, Hyungil; Jung, Yangil; Park, Dongjun; Park, Jeongyong; Kim, Ilhyun; Choi, Byungkwon; Jeong, Yonghwan; Park, Sangyoon

2013-04-01

This research was focused on the study for the manufacturing technology of HANA alloys to crease the corrosion resistance of 30% as well as the to improve the strength of 10% when compared to the commercial zirconium alloys. The new manufacturing concept having higher corrosion resistance and strength than commercial alloy performance can be obtained in this research. This result was transferred to the KNF and, that will be commercialized. This research result can be summarized like this; Ο Parameter study to increase formability of HANA alloy tube - Study on alloy element and heat-treatment effect - Study on texture development mechanism - Study on final annealing effect Ο Out-of-pile performance evaluation of HANA alloy tube - Corrosion performance evaluation of HANA alloy manufactured at KNF - Mechanical performance evaluation of HANA alloy manufactured at KNF - Recrystallization behavior evaluation of HANA alloy manufactured at KNF - Texture characterization of HANA alloy manufactured at KNF - Microstructure characterization of HANA alloy manufactured at KNF Ο Manufacturing guideline setup to increase formability of HANA alloy tube - Manufacturing guideline setup to decrease surface defect - Manufacturing guideline setup to increase strength and corrosion resistance - Manufacturing guideline setup to control texture

Investigation and basic evaluation for ultra-high burnup fuel cladding material

International Nuclear Information System (INIS)

Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

2001-03-01

In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

Zirconium cladding - the long way towards a mechanistic understanding of processing and performance

International Nuclear Information System (INIS)

Preuss, Michael

2011-01-01

Zirconium alloys are the material of choice to encapsulate nuclear fuel in light and heavy water-cooled reactors due to their low neutron absorption, excellent corrosion resistance and sufficient mechanical properties. Despite these advantageous physical and mechanical properties a more physically based understanding of microstructure and texture evolution during processing is highly desirable in order to improve our understanding of formability during thermomechanical processing and performance variability of cladding material. In addition, the purely empirical understanding of aqueous zirconium corrosion, hydrogen pick up, hydride precipitation as well as irradiation growth and creep limits the accuracy of life predictions and therefore the level of burnup that is obtained from current fuel assemblies. The presentation aims at giving examples of new research strategies that will enable the development of a new physical understanding of processing and performance aspects in zirconium cladding material, which is required to develop new predictive models. Particular emphasis will be placed on using novel research tools and large-scale research facilities such as neutron spallation and synchrotron radiation sources to undertake very detailed and often in-situ studies of deformation mechanisms and microstructure evolution as well as determining stress states in grain families, oxides and hydrides. The results will be presented in the view of how they might help us to improve our understanding and enable the development of better predictive models

Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

International Nuclear Information System (INIS)

Zmitko, M.

1999-01-01

In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

Water rod and fuel assembly

International Nuclear Information System (INIS)

Tsutsumi, Shinro; Tada, Nobuo; Nakajima, Junjiro; Aizawa, Yasuhiro.

1995-01-01

A water rod disposed in a fuel assembly comprises a larger diameter tube constituting an upwarding flow channel for coolants flown from the lower portion of a reactor core, and a smaller diameter tube connected fixedly to the larger diameter tube at the periphery of the upper end thereof and constituting a downwarding flow channel for coolants upwardly flown in the larger diameter tube. The larger diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to α + β treatment for once. The smaller diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to β treatment for once. With such procedures, the amount of irradiation growth of the tube in the axial direction is made greater in the larger diameter tube than that in the smaller diameter tube. Accordingly, since the smaller diameter tube is never bent by pressing, mechanical integrity of the fuel assembly is never lost. (I.N.)

Fuel performance update at Japanese BWRs

International Nuclear Information System (INIS)

Otsuka, Yasuyuki; Abe, Moriyasu

2008-01-01

Fuel Performance of Japanese BWRs has shown excellency with a very low fuel failure rate over the past couple of decades. In order to eliminate debris-fretting failures, which is considered as a dominant cause, efforts have been made to work out strict FME program and introduce lower tie plate with debris filters. Regarding a measure to detect failures without delay, an online off-gas monitor is installed, and there occurred no significant secondary failures leading to unnecessary plant shutdowns thanks to the monitor. Current standard fuel designs are 9x9 lattice configuration with the design maximum burnup set at 55 GWd/t and the bundle average discharge burnup at 45 GWd/t. Satisfactory performance of many fuels has gained so far to the burnup level. However recent data show an increase in hydrogen pickup of fuel claddings and spacers in the region of high burnup or high irradiation period, and this is one of the issues that should be addressed to maintain and improve fuel reliability for the reasons that hydrogen concentration may affect ductility of zirconium alloy. Furthermore, a recent study showed abnormally high hydrogen concentration in fuel cladding. Comprehensive root cause analyses were made, which include a poolside measurement for a number of high burnup fuels and development of new technique for tenacious crud removal, with the result that it is confirmed that a unique combination of cladding material, irradiation period and environment caused the phenomenon and it is not expected to occur again. New iron enhanced zirconium alloys, which have been investigated in for a long time, are expected to have greater resistance to hydrogen pickup, and introduction of the alloys is considered as the future materials. Regarding extensive inspections of fuel assemblies and channel boxes at Kashiwazaki-Kariwa NPS, no abnormality due to Niigata-ken Chuetsu-oki earthquake was observed for all plants. This paper deals with the current fuel performance at Japanese

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

Science.gov (United States)

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

Zirconium, calcium, and strontium contents in magnesium based biodegradable alloys modulate the efficiency of implant-induced osseointegration

Science.gov (United States)

Mushahary, Dolly; Sravanthi, Ragamouni; Li, Yuncang; Kumar, Mahesh J; Harishankar, Nemani; Hodgson, Peter D; Wen, Cuie; Pande, Gopal

2013-01-01

Development of new biodegradable implants and devices is necessary to meet the increasing needs of regenerative orthopedic procedures. An important consideration while formulating new implant materials is that they should physicochemically and biologically mimic bone-like properties. In earlier studies, we have developed and characterized magnesium based biodegradable alloys, in particular magnesium-zirconium (Mg-Zr) alloys. Here we have reported the biological properties of four Mg-Zr alloys containing different quantities of strontium or calcium. The alloys were implanted in small cavities made in femur bones of New Zealand White rabbits, and the quantitative and qualitative assessments of newly induced bone tissue were carried out. A total of 30 experimental animals, three for each implant type, were studied, and bone induction was assessed by histological, immunohistochemical and radiological methods; cavities in the femurs with no implants and observed for the same period of time were kept as controls. Our results showed that Mg-Zr alloys containing appropriate quantities of strontium were more efficient in inducing good quality mineralized bone than other alloys. Our results have been discussed in the context of physicochemical and biological properties of the alloys, and they could be very useful in determining the nature of future generations of biodegradable orthopedic implants. PMID:23976848

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Effect of zirconium addition on the ductility and toughness of cast zinc-aluminum alloy5, zamak5, grain refined by titanium plus boron

International Nuclear Information System (INIS)

Adnan, I.O.

2007-01-01

Zinc-aluminum casting alloys are frequently employed in design. They are inexpensive and have mechanical properties in many respects superior to aluminum and copper alloys. Common applications of zinc-aluminum alloys are in the automobile industry for manufacturing carburetors bodies, fuel pump bodies, driving wheels and door handles. They are mainly used for die casting due to their low melting points which ranges from 375 to 487 degree C, good fluidity, pollution free melting in addition to their high corrosion resistance. Against these advantages there exists the deficiency as these alloys solidify in a coarse dentititic structure which tends to deteriorate the mechanical properties and impact strength. It was found that addition of some rare earth materials e.g. titanium or titanium plus boron results in modifying its structure into a petal-like or nodular type. The available literature reveals that most of the published work is directed towards the metallurgical aspects and little or no work is published on the effect of those elements on its mechanical strength, ductility, toughness and impact strength. In this paper, the effect of addition of Zirconium on the microstructure, mechanical behavior, hardness, ductility and impact strength of zinc-aluminum alloy5, Zamak5, is investigated. It was found that addition of Ti+B or Zr or Ti+B+Zr resulted in modifying the coarse dentritic structure of the Zamak5 alloy into a fine nodular one. Further more, addition of any of these elements alone or together resulted in enhancement of the mechanical strength, hardness, ductility, toughness and impact strength of this alloy, for example an increase of 11% in hardness was achieved in case of Zr addition and 100% increase of ductility and 12.5% increase in impact strength were achieved in case of Ti+B addition. (author)

Corrosion Behavior and Oxide Properties of Zr-Nb-Cu and Zr-Nb-Sn Alloy in High Dissolved Hydrogen Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Lee, Yun Ju; Kim, Tae Ho; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-05-15

The water-metal interface is regarded as rate-controlling site governing the rapid oxidation transition in high burn-up fuel. And the zirconium oxide is made in water-metal interface and its structure and phase do an important role in terms of oxide properties. During oxidation process, the protective tetragonal oxide layer develops at the interface due to accumulated high stress during oxide growth, and it turns into non-protective monoclinic oxide with increasing oxide thickness, thus decreasing the stress. It has been reported that Nb addition was proven to be very beneficial for increasing the corrosion resistance of the zirconium alloys. From a more recent study, Cu addition in Nb containing Zirconium alloy was reported to be effective for increasing corrosion resistance in water containing B and Li. According to the previous research conducted, Zr-Nb-Cu shows better corrosion resistance than Zircaloy-4. The dissolved hydrogen (DH) concentration is the key issue of primary water chemistry, and the effect of DH concentration on the corrosion rate of nickel based alloy has been researched. However, the effect of DH on the zirconium alloy corrosion mechanism was not fully investigated. In this study, the weight gain measurement, FIB-SEM analysis, and Raman spectroscopic measurement were conducted to investigate the effects of dissolved hydrogen concentration and the chemical composition on the corrosion resistance and oxide phase of Zr-Nb-Cu alloy and Zr-Nb-Sn alloy after oxidizing in a primary water environment for 20 d. The corrosion rate of Zr-Nb-Cu alloy is slow, when it is compared to Zr-Nb-Sn alloy. In SEM images, the oxide thickness of Zr-Nb-Cu alloy is measured to be around 1.06 μm it of Zr-Nb-Sn alloy is measured to be 1.15 μm. It is because of the Segregation made by Sn solute element when Sn solute element oxidized. And according to ex situ Raman spectra, Zr-Nb-Cu alloy oxide has more tetragonal zirconium oxide fraction than Zr-Nb-Sn alloy oxide.

Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling

International Nuclear Information System (INIS)

Basin, N.

2001-01-01

This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl 4 + 2 Mg = 2 MgCl 2 . By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

Absorption of dissolved hydrogen from lithiated water during accelerated corrosion of zirconium-2.5 wt% niobium alloy

International Nuclear Information System (INIS)

Manolescu, A.V.; Mayer, P.; Rasile, E.M.; Mummenhoff, J.W.

1982-01-01

A series of laboratory experiments was carried out to determine the extent of dissolved hydrogen absorption from lithiated water by zirconium-2.5 wt% niobium alloy during corrosion. The material was exposed at 340 0 C to 1 M LiOH aqueous solution containing 0 to approximately 70 cm 3 /L of dissolved hydrogen. Results indicate that dissolved hydrogen has no effect on the corrosion rate or on the amount of hydrogen absorbed by the material

Contribution of in situ acoustic emission analysis coupled with thermogravimetry to study zirconium alloy oxidation

International Nuclear Information System (INIS)

Al Haj, O.; Peres, V.; Serris, E.; Cournil, M.; Grosjean, F.; Kittel, J.; Ropital, F.

2015-01-01

Zirconium alloy (zircaloy-4) corrosion behavior under oxidizing atmosphere at high temperature was studied using thermogravimetric experiment associated with acoustic emission analysis. Under a mixture of oxygen and air in helium, an acceleration of the corrosion is observed due to the detrimental effect of nitrogen which produces zirconium nitride. The kinetic rate increases significantly after a kinetic transition (breakaway). This acceleration is accompanied by an acoustic emission (AE) activity. Most of the acoustic emission bursts were recorded after the kinetic transition or during the cooling of the sample. Acoustic emission signals analysis allows us to distinguish different populations of cracks in the ZrO 2 layer. These cracks have also been observed by SEM on post mortem cross section of oxidized samples and by in-situ microscopy observations on the top surface of the sample during oxidation. The numerous small convoluted thin cracks observed deeper in the zirconia scale are not detected by the AE technique. From these studies we can conclude that mechanisms as irreversible mechanisms, as cracks initiation and propagation, generate AE signals

Hexagonal close packed to face centered cubic polymorphic transformation in nanocrystalline titanium-zirconium system by mechanical alloying

International Nuclear Information System (INIS)

Bera, S.; Manna, I.

2006-01-01

The present study reports a reversible hexagonal close packed (hcp) to face centered cubic (fcc) polymorphic phase transformation in four different nanocrystalline titanium-zirconium binary alloys in the course of mechanical alloying in a planetary ball mill. This transformation is monitored at appropriate stages by X-ray diffraction and high-resolution transmission electron microscopy. Lattice parameter of the nanocrystalline fcc phase is a function of the alloy composition. For a given alloy, the lattice parameter and hence volume per atom increase with increase in milling time under comparable conditions. On the other hand, crystallite size, measured from X-ray peak broadening, significantly decreases with the progress of milling. It is suggested that structural instability due to plastic strain, increasing lattice expansion, and negative (from core to boundary) hydrostatic pressure is responsible for this hcp → fcc polymorphic transformation. The said transformation seems reversible as isothermal annealing at 1000 deg. C for 1 h or melting the powder mass leads to partial or complete transformation of the milled product from single phase fcc to hcp

Immobilization of carbon 14 contained in spent fuel hulls through melting-solidification treatment

International Nuclear Information System (INIS)

Mizuno, T.; Maeda, T.; Nakayama, S.; Banba, T.

2004-01-01

The melting-solidification treatment of spent nuclear fuel hulls is a potential technique to improve immobilization/stabilization of carbon-14 which is mobile in the environment due to its weakly absorbing properties. Carbon-14 can be immobilized in a solid during the treatment under an inert gas atmosphere, where carbon is not oxidized to gaseous form and remains in the solid. A series of laboratory scale experiments on retention of carbon into an alloy waste form was conducted. Metallic zirconium was melted with metallic copper (Zr/Cu=8/2 in weight) at 1200 deg C under an argon atmosphere. Almost all of the carbon remained in the resulting zirconium-copper alloy. (authors)

Evaluation of the electrochemical behavior of U2.5Zr7.5Nb and U3Zr9Nb uranium alloys in relation to the pH and the solution aeration

International Nuclear Information System (INIS)

Mansur, Fabio Abud; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Figueiredo, Celia de Araujo

2011-01-01

The Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing, in cooperation with the Centro Tecnologico da Marinha (CTMSP), the advanced nuclear plate type fuel for the second core of the land-based reactor prototype of the Laboratorio de Geracao Nucleo-Eletrica (LABGENE). Recent investigations have shown that the fuel made of uranium-based niobium and zirconium alloys reaches the best performance relative to other fuels, e.g. UO 2 . Niobium and Zirconium also increase the corrosion resistance and the mechanical strength of the uranium alloys. By means of electrochemical techniques the corrosion behavior of alloys U 2 . 5 Zr 7.5 Nb and U 3 Zr 9 Nb, developed at CDTN and heat treated in the temperature range of 200 deg C to 600 deg C, was assessed. The effect of the parameters pH and solution aeration was studied as well as the influence of zirconium and niobium alloying elements in the corrosion of uranium. The techniques used were open circuit potential, electrochemical impedance and potentiodynamic anodic polarization at room temperature. The tests were performed in a three-electrode electrochemical cell with Ag/AgCl (3M KCl) as the reference electrode and a platinum plate as the auxiliary electrode. The potentiodynamic polarization curves of uranium and its alloys in acidic solutions showed regions with anodic currents limited by a passive film. The presence of niobium and zirconium contributed for the formation of this film. The impedance data showed the presence of two semicircles in the Bode diagram, indicating the occurrence of two distinct electrochemical processes. The data were fitted to an equivalent circuit model in order to obtain parameters of the electrochemical processes and evaluate the effect of the studied variables. (author)

Mobile Melt-Dilute Treatment for Russian Spent Nuclear Fuel

International Nuclear Information System (INIS)

Peacock, H.

2002-01-01

Treatment of spent Russian fuel using a Melt-Dilute (MD) process is proposed to consolidate fuel assemblies into a form that is proliferation resistant and provides critically safety under storage and disposal configurations. Russian fuel elements contain a variety of fuel meat and cladding materials. The Melt-Dilute treatment process was initially developed for aluminum-based fuels so additional development is needed for several cladding and fuel meat combinations in the Russian fuel inventory (e.g. zirconium-clad, uranium-zirconium alloy fuel). A Mobile Melt-Dilute facility (MMD) is being proposed for treatment of spent fuels at reactor site storage locations in Russia; thereby, avoiding the costs of building separate treatment facilities at each site and avoiding shipment of enriched fuel assemblies over the road. The MMD facility concept is based on laboratory tests conducted at the Savannah River Technology Center (SRTC), and modular pilot-scale facilities constructed at the Savannah River Site for treatment of US spent fuel. SRTC laboratory tests have shown the feasibility of operating a Melt-Dilute treatment process with either a closed system or a filtered off-gas system. The proposed Mobile Melt-Dilute process is presented in this paper

Uranium-zirconium based alloys part I: reference points for thermophysical properties

International Nuclear Information System (INIS)

Dias, Marcio Soares; Mattos, Joao Roberto L. de

2015-01-01

An integrated modelling process named Relative Variational Model (RVM) is in development by the fuel designers of the CDTN. The lack of measurements in the thermal and physical properties for new fuels, as well as the high dispersion of the existing measurements are challenges in the development of nuclear fuel concepts since that higher uncertainties of the material properties have as result the detrimental reduction on the safety margins . Based on the RVM, the integrated process has been applied to the derivation of reference points for the U-Zr based alloy. (author)

Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

International Nuclear Information System (INIS)

Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

1997-01-01

The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

Oxidation behaviour of Zr-Ce alloys. Kinetic and microstructure aspects

International Nuclear Information System (INIS)

Rouillon, Ludovic

1996-01-01

As Zircaloy alloys are used for fuel rods in pressurized water nuclear reactors, this research thesis aims at studying and improving corrosion resistance of zirconium alloys while maintaining their mechanical properties. It more precisely deals with the kinetic and microstructure aspects of the external corrosion of the cladding by the coolant. In the case of Zircaloys, this corrosion is characterized by a kinetic transition from an initially parabolic to a linear regime. This research aims at intervening on this transition by elaborating zirconium alloys containing an element which stabilizes zirconia, in this case cerium. After having reported a bibliographical study on sheath oxidation, on parameters which influence sheath oxidation kinetics, on zirconia stabilization by doping elements, on the interest of lanthanide oxides, the author reports a feasibility study on the use of cerium (choice and preparation, sintered ceramic characterization, annealing of stabilized zirconia), reports a metallurgical study of Zr-Ce alloys, reports the study of the oxidation behaviour of these alloys (in autoclave, in presence of oxygen, under oxygen and then water) and the characterization of the microstructures of the oxide layers. He finally discusses the relationship between microstructure and oxidation kinetics, the role of cerium in the oxidation process, and the role of water in the oxidation process [fr

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Oxidation kinetics of some zirconium alloys in flowing carbon dioxide at high temperatures

International Nuclear Information System (INIS)

Kohli, R.

1980-01-01

The oxidation kinetics of three zirconium alloys (Zr-2.2 wt% Hf, Zr-2.5 wt% Nb, and Zr-3 wt% Nb-1 wt% Sn) have been measured in flowing carbon dioxide in the temperature range from 873 to 1173 K to 120 ks (2000 min). At all oxidation temperatures, Zr-2.5 Nb and Zr-3 Nb-1 Sn showed a transition to rapid linear kinetics after initial parabolic oxidation. The Zr-2.2 Hf showed this transition at temperatures in the range from 973 to 1173 K; at 873 K, no transition was observed within the oxidation times reported. The Zr-2.2 Hf showed the smallest weight gains, followed in order by Zr-2.5 Nb and Zr-3 Nb-1 Sn. Increased oxidation rates and shorter times-to-rate-transition of Zr-2.2 Nb and Zr-1 Sn as compared with Zr-2.2 Hf can be attributed to the presence of niobium, tin, and hafnium in the alloys. This is considered in terms of the Nomura-Akutsu model, according to which hafnium should delay the rate transition, while niobium and tin lead to shorter times-to-rate-transition. The scale on Zr-2.2 Hf was identified as monoclinic zirconia, while the tetragonal phase, 6ZrO 2 .Nb 2 O 5 , was contained in the monoclinic zirconia scales on both other alloys

Corrosion performance of new Zircaloy-2-based alloys

International Nuclear Information System (INIS)

Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

1994-01-01

A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

Atomistic modeling of zirconium hydride precipitation: methodology for deriving a tight-binding potential

International Nuclear Information System (INIS)

Dufresne, Alice

2014-01-01

The zirconium-hydrogen system is of nuclear safety interest, as the hydride precipitation leads to the cladding embrittlement, which is made of zirconium-based alloys. The cladding is the first safety barrier confining the radioactive products: its integrity shall be kept during the entire fuel-assemblies life, in reactor, including accidental situation, and post-operation (transport and storage). Many uncertainties remain regarding the hydrides precipitation kinetics and the local stress impact on their precipitation. The atomic scale modeling of this system would bring clarifications on the relevant mechanisms. The usual atomistic modeling methods are based on thermo-statistic approaches, whose precision and reliability depend on the interatomic potential used. However, there was no potential allowing a rigorous study of the Zr-H system. The present work has indeed addressed this issue: a new tight-binding potential for zirconium hydrides modeling is now available. Moreover, this thesis provides a detailed manual for deriving such potentials accounting for spd hybridization, and fitted here on DFT results. This guidebook has be written in light of modeling a pure transition metal followed by a metal-covalent coupling (metallic carbides, nitrides and silicides). (author)

A computer model for hydride blister growth in zirconium alloys

International Nuclear Information System (INIS)

White, A.J.; Sawatzky, A.; Woo, C.H.

1985-06-01

The failure of a Zircaloy-2 pressure tube in the Pickering unit 2 reactor started at a series of zirconium hydride blisters on the outside of the pressure tube. These blisters resulted from the thermal diffusion of hydrogen to the cooler regions of the pressure tube. In this report the physics of thermal diffusion of hydrogen in zirconium is reviewed and a computer model for blister growth in two-dimensional Cartesian geometry is described. The model is used to show that the blister-growth rate in a two-phase zirconium/zirconium-hydride region does not depend on the initial hydrogen concentration nor on the hydrogen pick-up rate, and that for a fixed far-field temperature there is an optimum pressure-type/calandria-tube contact temperature for growing blisters. The model described here can also be used to study large-scale effects, such as hydrogen-depletion zones around hydride blisters

Study on direct dissolution of U-10Zr alloy and distribution of uranium and zirconium in liquid cadmium

International Nuclear Information System (INIS)

Ye Yuxing; Gao Yuan

1997-09-01

The effect of dissolution time, temperature, total surface area of U-10Zr alloy pellets and stirring on the dissolution and dissolution rate of uranium in liquid cadmium were studied. Cadmium containing U and Zr dissolved from U-10Zr alloy at 475 degree C and 500 degree C respectively was analyzed with electron microanalyzer. The experimental results show that at 400 degree and 500 degree C with the stirring rate of some 150 r/min, the solubilities of uranium in liquid cadmium are 0.4% and 2.2%, respectively. At the first 30 min, the dissolution rates of U-10Zr alloy pellets are 0.05 g/(cm 2 ·h) and 0.32 g/(cm 2 ·h), respectively. The suitable dissolution conditions for U-10Zr alloy pellets in liquid cadmium (the ratio of the mass of liquid cadmium to that of the pellets ≅7) are: temperature, about 480 degree C; stirring rate, about 150 r/min; dissolution time, 4 h. The distribution of uranium and zirconium in cadmium is homogeneous

Study on the improvement of the properties of Zr alloys

International Nuclear Information System (INIS)

Kim, Young Suk; Han, Jung Ho; Jeong, Yong Hwan; Lee, Duk Hyun; Park, Gi Sung; Hong, Jun Hwa; Park, Ji Yun; No, Gae Ho

1992-01-01

1) The objective of this study is to develop the corrosion resistant zirconium base alloys. In order to achieve this goal, this year's activities have focused on the guidelines for the corrosion resistant zirconium alloy design, the manufacturing of the sheets of zirconium base alloys and finally the characterization of the NAZAs (New Alternate Zirconium alloys). The main results from this study can be summarized as follows: 2) Based on the evaluation of the role of alloying elements, i.e., Nb, Sn, Fe, Cr, and etc, as many as 23 different kinds of the NAZAs were preliminarily designed. 3) The 3 kinds of the NAZAs-Lot 15, 22 and 23 manufactured into a sheet though a series of manufacturing procedures. 4) The microstructures, hardness and the corrosion performances of 3 kinds of NAZAs were investigated. (Author)

Progress in developing very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

1999-01-01

Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

Evaluation of the electrochemical behavior of U{sub 2.5}Zr{sub 7.5}Nb and U{sub 3}Zr{sub 9}Nb uranium alloys in relation to the pH and the solution aeration

Energy Technology Data Exchange (ETDEWEB)

Mansur, Fabio Abud; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Figueiredo, Celia de Araujo, E-mail: ferraz@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing, in cooperation with the Centro Tecnologico da Marinha (CTMSP), the advanced nuclear plate type fuel for the second core of the land-based reactor prototype of the Laboratorio de Geracao Nucleo-Eletrica (LABGENE). Recent investigations have shown that the fuel made of uranium-based niobium and zirconium alloys reaches the best performance relative to other fuels, e.g. UO{sub 2}. Niobium and Zirconium also increase the corrosion resistance and the mechanical strength of the uranium alloys. By means of electrochemical techniques the corrosion behavior of alloys U{sub 2}.{sub 5}Zr{sub 7.5}Nb and U{sub 3}Zr{sub 9}Nb, developed at CDTN and heat treated in the temperature range of 200 deg C to 600 deg C, was assessed. The effect of the parameters pH and solution aeration was studied as well as the influence of zirconium and niobium alloying elements in the corrosion of uranium. The techniques used were open circuit potential, electrochemical impedance and potentiodynamic anodic polarization at room temperature. The tests were performed in a three-electrode electrochemical cell with Ag/AgCl (3M KCl) as the reference electrode and a platinum plate as the auxiliary electrode. The potentiodynamic polarization curves of uranium and its alloys in acidic solutions showed regions with anodic currents limited by a passive film. The presence of niobium and zirconium contributed for the formation of this film. The impedance data showed the presence of two semicircles in the Bode diagram, indicating the occurrence of two distinct electrochemical processes. The data were fitted to an equivalent circuit model in order to obtain parameters of the electrochemical processes and evaluate the effect of the studied variables. (author)

Gel structure of the corrosion layer on cladding pipes of nuclear fuels

Czech Academy of Sciences Publication Activity Database

Medek, Jiří; Weishauptová, Zuzana

2009-01-01

Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

Structural studies of calcium phosphate doped with titanium and zirconium obtained by high-energy mechanical alloying

Energy Technology Data Exchange (ETDEWEB)

Silva, C C; Sombra, A S B [Telecommunications and Materials Science and Engineering Laboratory (LOCEM), Physics Department, Federal University of Ceara, Campus do Pii, Postal Code 6030, 60455-760, Fortaleza-Ceara (Brazil)], E-mail: sombra@fisica.ufc.br

2009-12-15

In this paper, we present a new variation of the solid-state procedure on the synthesis of bioceramics with titanium (CapTi) and zirconium (CapZr), considering that zirconium (ZrO{sub 2}) and titanium oxide (TiO{sub 2}) are strengthening agents, due to their superb force and fracture toughness. The high efficiency of the calcination process opens a new way of producing commercial amounts of nanocrystalline bioceramics. In this work, a new variation of the solid-state procedure method was used to produce nanocrystalline powders of titanium and zirconium, using two different experimental chemical routes: CapTi: Ca(H{sub 2}PO{sub 4}){sub 2}+TiO{sub 2} and CapZr: Ca(H{sub 2}PO{sub 4}){sub 2}+ZrO{sub 2}. The powders were submitted to calcination processes (CapTic and CapZrc) at 800, 900 and 1000 deg. C. The calcium titanium phosphate phase, CaTi{sub 4}P{sub 6}O{sub 24}, was obtained in the CapTic reaction and the calcium zirconium phosphate, CaZr{sub 4}P{sub 6}O{sub 24}, was obtained in the CapZrc reaction. The obtained ceramics were characterized by x-ray powder diffraction (XRD), infrared (IR) spectroscopy, Raman scattering spectroscopy (RSS) and scanning electron microscopy (SEM) analysis. This method was compared with the milling process (CapTim and CapZrm), where in the last process the melting is not necessary and the powder obtained is nanocrystalline. The calcium titanium phosphate phase, CaTi{sub 4}P{sub 6}O{sub 24}, was obtained in the reaction CapTim, but in CapZrm the formation of any calcium phosphate phase even after 15 h of dry mechanical alloying was not observed.

Determination of plastic anisotropy of zirconium alloys cladding

International Nuclear Information System (INIS)

Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

1991-01-01

Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

Third international spent fuel storage technology symposium/workshop: proceedings. Volume 2

International Nuclear Information System (INIS)

1986-01-01

The scope of this meeting comprised dry storage and rod consolidation, emphasizing programs on water reactor fuel with zirconium alloy cladding. Volume 2 contains the papers from the poster session and workshops that were conducted during the meeting. There were 18 poster presentations. Four workshops were held: Fuel Integrity; Storage System Modeling and Analysis; Rod Consolidation Technology; and System Integration and Optimization. Individual papers were processed for inclusion in the Energy Data Base

Method to precipitate metals on zirconium articles without using current

International Nuclear Information System (INIS)

Donaghy, R.E.

1978-01-01

Tubes or other bodies made of zirconium or Zr alloys used for taking up nuclear fuels are coated without current with a metal film to improve the mechanical and corrosion properties. The article is activated in a solution of ammonium bifluoride and sulphuric acid where an electrical conducting fixed and loose layer are fermed. This loose film is removed chemically by reacting with fluoroboric acid or hydrofluorosilica acid by ultrasonic treatment or by stripping from organic material (cotton, polyester, nylon). The current-free plating with Cu and Ni is described. The article is washed with deionized water between the process steps and finally degassed at a temperature of 150-200 0 C. (IHOE) [de

Susceptibility of cold-worked zirconium-2.5 wt% niobium alloy to delayed hydrogen cracking

International Nuclear Information System (INIS)

Coleman, C.E.

1976-01-01

Notched tensile specimens of cold-worked zirconium-2.5 wt% niobium alloy have been stressed at 350 K and 520 K. At 350 K, above a possible threshold stress of 200 MPa, specimens exhibited delayed failure which was attributed to hydride cracking. Metallography showed that hydrides accumulated at notches and tips of growing cracks. The time to failure appeared to be independent of hydrogen content over the range 7 to 100 ppm hydrogen. Crack growth rates of about 10 -10 m/s deduced from fractography were in the same range as those necessary to fracture pressure tubes. The asymptotic stress intensity for delayed failure, Ksub(1H), appeared to be about 5 MPa√m. With this low value of Ksub(1H) small surface flaws may propagate in pressure tubes which contain large residual stresses. Stress relieving and modified rolling procedures will reduce the residual stresses to such an extent that only flaws 12% of the wall thickness or greater will grow. At 520 K no failures were observed at times a factor of three greater than times to failure at 350 K. Zirconium-2.5 wt% niobium appears to be safe from delayed hydrogen cracking at the reactor operating temperature. (author)

Modelling zirconium hydrides using the special quasirandom structure approach

KAUST Repository

Wang, Hao; Chroneos, Alexander I.; Jiang, Chao; Schwingenschlö gl, Udo

2013-01-01

The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH2-x (x = 0.25-0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys. © the Owner Societies 2013.

Corrosion behaviour of zirconium alloys in the autoclaves of Embalse nuclear power plant

International Nuclear Information System (INIS)

Bordoni, Roberto A.; Olmedo, Ana M.; Villegas, Marina; Miyagusuku, Marcela; Maroto, Alberto J. G.; Sainz, Ricardo A.; Fernandez, Alberto N.; Allemandi, Walter D.

1999-01-01

The corrosion behaviour of zirconium alloys coupons attached to the holders of the autoclaves located out of core in the primary circuit of Embalse nuclear power plant is described. The Zr-2.5 Nb coupons of the autoclaves at the higher temperature (305 C degrees) and the Zry-4 coupons of the autoclaves at 265 and 305 C degrees installed in 1988 had a normal corrosion behaviour, after 3500 of full power days. While, the Zr-2.5 Nb coupons, at 265 C degrees, showed the presence of white oxide nuclei and a weight gain indicating an abnormal corrosion behaviour which might be attributed to the material microstructure. Complementary tests, made in the period September 1991-April 1993, showed that the abnormal corrosion behaviour observed for the Canadian coupons installed in 1983 was due to a surface contamination of the Zry-4 coupons and due to the microstructure of the Zr-2.5 Nb coupons. The normal corrosion behaviour for both alloys installed in 1986, showed that the resin ingress to the primary circuit that occurred in 1988, do not affect the performance of these materials. (author)

Lithium uptake and the corrosion of zirconium alloys in aqueous lithium hydroxide solutions

International Nuclear Information System (INIS)

Ramasubramanian, N.

1991-01-01

This paper reports on corrosion films on zirconium alloys that were analyzed for lithium by Atomic Absorption Spectroscopy (AAS), Secondary Ion Mass Spectrometry (SIMS), and Infrared Reflection Absorption Spectroscopy (IRAS). The oxides grown in reactor in dilute lithium hydroxide solution, specimens cut from Zircaloy, and Zr-2.5Nb alloy pressure tubes removed from CANDU (Canada Deuterium Uranium, Registered Trademark) reactors showed low concentrations of lithium (4 to 50 ppm). The lithium was not leachable in a warm dilute acid. 6 Li undergoes transmutation by the 6 Li(n,t) 4 He reaction. However, SIMS profiles for d 7 Li were identical through the bulk oxide and the isotopic ratio was close to the natural abundance value. The lithium in the oxide, existing as adsorbed lithium on the surface, has been in dynamic equilibrium with lithium in the coolant, and, in spite of many Effective Full Power Years (EFPY) of operation, lithium added to the CANDU coolant at ∼2.5 ppm is not concentrating in the oxides. On the other hand, corrosion films grown in the laboratory in concentrated lithium hydroxide solutions were very porous and contained hundreds of ppm of lithium in the oxide

Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

International Nuclear Information System (INIS)

Rico, A.; Martin-Rengel, M.A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F.J.

2014-01-01

It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found

U-Zr-RE Fuel Alloy with Minor Actinides

Energy Technology Data Exchange (ETDEWEB)