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

Chemical dissolution of spent fuel and cladding using complexed fluoride species

International Nuclear Information System (INIS)

Rance, P.J.W.; Freeman, G.A.; Mishin, V.; Issoupov, V.

2001-01-01

The dissolution of LWR fuel cladding using two fluoride ion donors, HBF 4 and K 2 ZrF 6 , in combination with nitric acid has been investigated as a potential reprocessing head-end process suitable for chemical decladding and fuel dissolution in a single process step. Maximum zirconium concentrations in the order of 0,75 to 1 molar have been achieved and dissolution found to continue to low F:Zr ratios albeit at ever decreasing rates. Dissolution rates of un-oxidised zirconium based fuel claddings are fast, whereas oxidised materials exhibit an induction period prior to dissolution. Data is presented relating to the rates of dissolution of cladding and UO 2 fuels under various conditions. (author)

Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

International Nuclear Information System (INIS)

Cabrera-Salcedo, A.

2012-01-01

During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

Solid-phase zirconium and fluoride species in alkaline zircaloy cladding waste at Hanford.

Science.gov (United States)

Reynolds, Jacob G; Huber, Heinz J; Cooke, Gary A; Pestovich, John A

2014-08-15

The United States Department of Energy Hanford Site, near Richland, Washington, USA, processed plutonium between 1944 and 1987. Fifty-six million gallons of waste of various origins remain, including waste from removing zircaloy fuel cladding using the so-called Zirflex process. The speciation of zirconium and fluoride in this waste is important because of the corrosivity and reactivity of fluoride as well as the (potentially) high density of Zr-phases. This study evaluates the solid-phase speciation of zirconium and fluoride using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Two waste samples were analyzed: one waste sample that is relatively pure zirconium cladding waste from tank 241-AW-105 and another that is a blend of zirconium cladding wastes and other high-level wastes from tank 241-C-104. Villiaumite (NaF) was found to be the dominant fluoride species in the cladding waste and natrophosphate (Na7F[PO4]2 · 19H2O) was the dominant species in the blended waste. Most zirconium was present as a sub-micron amorphous Na-Zr-O phase in the cladding waste and a Na-Al-Zr-O phase in the blended waste. Some zirconium was present in both tanks as either rounded or elongated crystalline needles of Na-bearing ZrO2 that are up to 200 μm in length. These results provide waste process planners the speciation data needed to develop disposal processes for this waste. Copyright © 2014 Elsevier B.V. All rights reserved.

Development of advanced zirconium fuel cladding

International Nuclear Information System (INIS)

Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

2007-04-01

This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

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.

Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

1986-09-01

Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

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.

Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

International Nuclear Information System (INIS)

Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun

2015-01-01

The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings

Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings.

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)

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...

WWER water chemistry related to fuel cladding behaviour

Energy Technology Data Exchange (ETDEWEB)

Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

1997-02-01

Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

International Nuclear Information System (INIS)

Chatterjee, S.

2005-01-01

During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

Manufacturing process for the metal ceramic hybrid fuel cladding tube

International Nuclear Information System (INIS)

Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

2012-01-01

For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

Manufacturing process for the metal ceramic hybrid fuel cladding tube

Energy Technology Data Exchange (ETDEWEB)

Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

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)

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.)

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

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.

Out-of-pile test of zirconium cladding simulating reactivity initiated accident

Energy Technology Data Exchange (ETDEWEB)

Kim, J. H.; Lee, M. H.; Choi, B. K.; Bang, J. K.; Jung, Y. H. [KAERI, Taejon (Korea, Republic of)

2004-07-01

Mechanical properties of zirconium cladding such as Zircaloy-4 and advanced cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) as an out-pile test. Cladding was hydrided by means of charging hydrogen up to 1000ppm to simulate high-burnup situation, finally fabricated to circumferential tensile specimen. Ring tension test was carried out from 0.01 to 1/sec to keep pace with actual RIA event. The results showed that mechanical strength of zirconium cladding increased at the value of 7.8% but ductility decreased at the 34% as applied strain rate and absorbed hydrogen increased. Further activities regarding out-of-pile testing plans for simulated high-burnup cladding were discussed in this paper.

High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

International Nuclear Information System (INIS)

Hu, J.; Lozano-Perez, S.; Grovenor, C.

2015-01-01

Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

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

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

Energy Technology Data Exchange (ETDEWEB)

Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-02-01

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

International Nuclear Information System (INIS)

Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

2016-01-01

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

Method for automatic filling of nuclear fuel rod cladding tubes

International Nuclear Information System (INIS)

Bezold, H.

1979-01-01

Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

Zirconium alloy barrier having improved corrosion resistance

International Nuclear Information System (INIS)

Adamson, R.B.; Rosenbaum, H.S.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

Evolutionary developments of advanced PWR nuclear fuels and cladding materials

International Nuclear Information System (INIS)

Kim, Kyu-Tae

2013-01-01

Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

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

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)

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.

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)

Mechanism for iodine cracking of zirconium claddings

International Nuclear Information System (INIS)

Novikov, V.V.

1991-01-01

The mechanism of iodine cracking of zirconium cladding is analyzed taking into account the effect of stresses on diffusion. A decisive effect of the stress gradiemt on crack propagation in an agressive medium is shown. The experimental data are compared with the proposed model

Separate-effect tests on zirconium cladding degradation in air ingress situations

Energy Technology Data Exchange (ETDEWEB)

Duriez, C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Steinbrueck, M. [Forschungszentrum Karlsruhe, FZK, Institut fuer Materialforschung, Postfach 3640, 76021 Karlsruhe (Germany); Ohai, D.; Meleg, T. [Institute for Nuclear Research, INR, Nuclear Material and Corrosion Department, Pitesti, 115400 Mioveni Arges (Romania); Birchley, J.; Haste, T. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

2009-02-15

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of

Separate-effect tests on zirconium cladding degradation in air ingress situations

International Nuclear Information System (INIS)

Duriez, C.; Steinbrueck, M.; Ohai, D.; Meleg, T.; Birchley, J.; Haste, T.

2009-01-01

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale

DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

International Nuclear Information System (INIS)

Rudisill, T; John Mickalonis, J

2006-01-01

The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

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)

About criteria of inadmissible embrittlement of zirconium fuel cladding during LOCA in the PWRs

International Nuclear Information System (INIS)

Osmachkin, V.S.

1999-01-01

According the licensing procedures the designers of the PWRs reactor have to prove the meeting of special safety requirements. One criteria on effectiveness of the Emergency Core Cooling System is not to exceeding some limited conditions of the fuel cladding during LOCA accidents (typical example T m ax o C, ECR<0,17 and oth.). The damage of fuel element in the core during LOCA is caused by the oxidation of the cladding, its embrittlement and thermal shock stresses after initiation of the heat removal by a cold water from emergency core cooling system. In the paper the conservatism in criteria to avoid brittle ruptures of the fuel elements is discussed. Taking into account the influence of fuel burnup on the property of the cladding and a potential presence of air in the steam, it is believed that criteria of survivability of the zircaloy fuel cladding during LOCA may not be enough conservative.(author)

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

Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

International Nuclear Information System (INIS)

Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

2009-01-01

To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

Cladding failure model III (CFM III). A simple model for iodine induced stress corrosion cracking of zirconium-lined barrier and standard zircaloy cladding

International Nuclear Information System (INIS)

Tasooji, A.; Miller, A.K.

1984-01-01

A previously developed unified model (SCCIG*) for predicting iodine induced SCC in standard Zircaloy cladding was modified recently into the ''SCCIG-B'' model which predicts the stress corrosion cracking behaviour of zirconium lined barrier cladding. Several published papers have presented the capability of these models for predicting various observed behaviours related to SCC. A closed form equation, called Cladding Failure Model III (CMFIII), has been derived from the SCCIG-B model. CFMIII takes the form of an explicit equation for the radial crack growth rate dc/dt as a function of hoop strain, crack depth, temperature, and surface iodine concentration in irradiated cladding (both barrier and standard Zircaloy). CMFIII has approximately the same predictive capabilities as the physically based SCCIG and/or SCCIG-B models but is computationally faster and more convenient and can be easily utilized in fuel performance codes for predicting the behaviour of barrier and standard claddings in reactor operations. (author)

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

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.)

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)

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

Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

International Nuclear Information System (INIS)

Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

1986-01-01

Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

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

Analyses on Silicide Coating for LOCA Resistant Cladding

Energy Technology Data Exchange (ETDEWEB)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2015-10-15

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

Analyses on Silicide Coating for LOCA Resistant Cladding

International Nuclear Information System (INIS)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

2015-01-01

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

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

Investigation and recovery of unrecovered fuel pellets and cladding tube pieces

International Nuclear Information System (INIS)

Kobayashi, Keiji

1980-01-01

The total weight of the fuel pellets lost due to break was about 1206 g, and cladding tube pieces were about 217 g. Among these, the pellets of about 527 g and the cladding tube pieces of about 152 g were recovered when broken fuel rods were discovered. It is not desirable to leave these broken pieces as unrecovered in view of safety and the management of nuclear fuel materials. Kansai Electric Power Co., Inc., investigated the position and the amount of these pellets and cladding tube pieces for about a year, and recovered a part of them. The results were written in two reports. The objects of the investigation and recovery, and the method of recovery are explained. The UO 2 and zirconium recovered were 58.52 g and 369.58 g, respectively. The solid pellets were recovered from the reactor, fuel assemblies, a spent fuel pit and canals, and the content in sludge was recovered from other installations. The amounts of unrecovered pellets and cladding tube pieces in primary cooling water, coolant filters, sealing water filters, primary cooling pipes, waste resins and fuel assemblies were estimated. The problems concerning the recovery and estimation are pointed out. The results of estimating the amount of uranium in coolant filters and sealing water filters are useful to know the time of the occurrence of accident. (Kako, I.)

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

Energy Technology Data Exchange (ETDEWEB)

Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

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

Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

Energy Technology Data Exchange (ETDEWEB)

Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

2007-02-15

Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

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-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

Science.gov (United States)

Mariani, Robert Dominick

2014-09-09

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

Energy Technology Data Exchange (ETDEWEB)

Rosenbaum, H.S. (comp.)

1980-09-01

Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1980-09-01

Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U

Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Fourth semiannual report, July-December 1980

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1981-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts have been developed for possible demonstration: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the scope of this program one of these concepts had to be selected for a large-scale demonstration in a commercial power reactor. The selection was made to demonstrate Zr-liner fuel and to include bundles which have liners prepared from either low oxygen sponge zirconium or of crystal bar zirconium. The demonstration is intended to include a total of 132 barrier bundles in the reload for Quad Cities Unit 2, Cycle 6. In the current report period changes in the nuclear design were made to respond to changes in the Energy Utilization Plan for Quad Cities Unit 2. Bundle designs were completed, and were licensed for use in a BWR/3. The core specific licensing will be done as part of the reload license for Quad Cities Unit 2, Cycle 6

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.)

Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

International Nuclear Information System (INIS)

Leiling, W.

1981-12-01

This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

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.)

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.

Study of the uranium-zirconium diffusion; Etude de la diffusion uranium-zirconium

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mairy, C; Bouchet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The intermetallic diffusion of uranium fuel and zirconium used as cladding is studied. Intermetallic diffusion can occur during the cladding of uranium rods and uranium can penetrate the zirconium cladding. Different parameters are involved in this mechanism as structure and mechanical properties of the diffusion area as well as presence of impurities in the metal. The uses of different analysis techniques (micrography, Castaing electronic microprobe, microhardness and autoradiography) have permitted to determine with great accuracy the diffusion coefficient in gamma phase (body centered cubic system) and the results have given important information on the intermetallic diffusion mechanisms. The existence of the Kirkendall effect in the U-Zr diffusion is also an argument in favor of the generality of the diffusion mechanism by vacancies in body centered cubic system. (M.P.)

FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [ORNL

2014-03-28

The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

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.)

Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar; Allen, Todd; Cole, James

2013-02-27

The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

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.

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

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

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

A deformation and thermodynamic model for hydride precipitation kinetics in spent fuel cladding

International Nuclear Information System (INIS)

Stout, R.B.

1989-10-01

Hydrogen is contained in the Zircaloy cladding of spent fuel rods from nuclear reactors. All the spent fuel rods placed in a nuclear waste repository will have a temperature history that decreases toward ambient; and as a result, most all of the hydrogen in the Zircaloy will eventually precipitate as zirconium hydride platelets. A model for the density of hydride platelets is a necessary sub-part for predicting Zircaloy cladding failure rate in a nuclear waste repository. A model is developed to describe statistically the hydride platelet density, and the density function includes the orientation as a physical attribute. The model applies concepts from statistical mechanics to derive probable deformation and thermodynamic functionals for cladding material response that depend explicitly on the hydride platelet density function. From this model, hydride precipitation kinetics depend on a thermodynamic potential for hydride density change and on the inner product of a stress tensor and a tensor measure for the incremental volume change due to hydride platelets. The development of a failure response model for Zircaloy cladding exposed to the expected conditions in a nuclear waste repository is supported by the US DOE Yucca Mountain Project. 19 refs., 3 figs

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

Allowable peak heat-up cladding temperature for spent fuel integrity during interim-dry storage

Directory of Open Access Journals (Sweden)

Ki-Nam Jang

2017-12-01

Full Text Available To investigate allowable peak cladding temperature and hoop stress for maintenance of cladding integrity during interim-dry storage and subsequent transport, zirconium alloy cladding tubes were hydrogen-charged to generate 250 ppm and 500 ppm hydrogen contents, simulating spent nuclear fuel degradation. The hydrogen-charged specimens were heated to four peak temperatures of 250°C, 300°C, 350°C, and 400°C, and then cooled to room temperature at cooling rates of 0.3 °C/min under three tensile hoop stresses of 80 MPa, 100 MPa, and 120 MPa. The cool-down specimens showed that high peak heat-up temperature led to lower hydrogen content and that larger tensile hoop stress generated larger radial hydride fraction and consequently lower plastic elongation. Based on these out-of-pile cladding tube test results only, it may be said that peak cladding temperature should be limited to a level < 250°C, regardless of the cladding hoop stress, to ensure cladding integrity during interim-dry storage and subsequent transport.

Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

International Nuclear Information System (INIS)

Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

2002-01-01

Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

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

Fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

2015-01-01

The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

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

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

BWR fuel clad behaviour following LOCA

International Nuclear Information System (INIS)

Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

1996-01-01

Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

International Nuclear Information System (INIS)

Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

2015-01-01

This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

Current status of materials development of nuclear fuel cladding tubes for light water reactors

International Nuclear Information System (INIS)

Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

2017-01-01

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

Energy Technology Data Exchange (ETDEWEB)

Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

2017-05-15

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

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.)

Titanium(IV), zirconium, hafnium and thorium

International Nuclear Information System (INIS)

Brown, Paul L.; Ekberg, Christian

2016-01-01

Titanium can exist in solution in a number of oxidation states. The titanium(IV) exists in acidic solutions as the oxo-cation, TiO 2+ , rather than Ti 4+ . Zirconium is used in the ceramics industry and in nuclear industry as a cladding material in reactors where its reactivity towards hydrolysis reactions and precipitation of oxides may result in degradation of the cladding. In nature, hafnium is found together with zirconium and as a consequence of the contraction in ionic radii that occurs due to the 4f -electron shell, the ionic radius of hafnium is almost identical to that of zirconium. All isotopes of thorium are radioactive and, as a consequence of it being fertile, thorium is important in the nuclear fuel cycle. The polymeric hydrolysis species that have been reported for thorium are somewhat different to those identified for zirconium and hafnium, although thorium does form the Th 4 (OH) 8 8+ species.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

Energy Technology Data Exchange (ETDEWEB)

Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

1965-06-28

Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

Reuse of spent fuel cladding Zr by molten salt toward advanced recycle society

International Nuclear Information System (INIS)

Amano, Osamu; Kobayashi, Hiroaki; Suzuki, Kazunori; Yasuike, Y.; Sato, Nobuaki

2003-01-01

Cladding tubes of zircaloy 95% generated from reprocessing process for spent nuclear fuels are to be chopped in about 3 cm length, compacted and solidified with cements. This paper reports the summary of investigation of the present possible techniques for zirconium recovery: (1) electrolysis of molten salts (Zr-chlorides and/or fluorides) and (2) separation as volatile zirconium chlorides (ZrCl 4 ) (chloride volatility process) followed by reaction with metallic magnesium at 900degC to produce sponged Zr (Kroll method). The feasibility are discussed from the point of view of reduction of secondary radioactive wastes, accumulation of such nuclides as Co-60 and Ni-63 in electrolytic basin, radioactivity estimation in the products, and also problems of cleaning and reducing chemicals. (S. Ohno)

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

Analysis of hafnium in zirconium alloys

International Nuclear Information System (INIS)

Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

1977-01-01

It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

International Nuclear Information System (INIS)

Chosson, Raphael

2014-01-01

During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

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.

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

International Nuclear Information System (INIS)

Rose, K.M.; Mann, C.A.; Hindle, E.D.

1979-01-01

In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Rose, K.M.; Mann, C.A.; Hindle, E.D.

1979-12-01

In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

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)

Thermodynamics of pellet-cladding interaction

International Nuclear Information System (INIS)

Kyoh, Bunkei; Fuji, Kensho

1987-01-01

Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

2014-08-15

Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings

Science.gov (United States)

Yeom, Hwasung; Lockhart, Cody; Mariani, Robert; Xu, Peng; Corradini, Michael; Sridharan, Kumar

2018-02-01

This study investigates steam corrosion of bulk ZrSi2, pure Si, and zirconium-silicide coatings as well as water quenching behavior of ZrSi2 coatings to evaluate its feasibility as a potential accident-tolerant fuel cladding coating material in light water nuclear reactor. The ZrSi2 coating and Zr2Si-ZrSi2 coating were deposited on Zircaloy-4 flats, SiC flats, and cylindrical Zircaloy-4 rodlets using magnetron sputter deposition. Bulk ZrSi2 and pure Si samples showed weight loss after the corrosion test in pure steam at 400 °C and 10.3 MPa for 72 h. Silicon depletion on the ZrSi2 surface during the steam test was related to the surface recession observed in the silicon samples. ZrSi2 coating (∼3.9 μm) pre-oxidized in 700 °C air prevented substrate oxidation but thin porous ZrO2 formed on the coating. The only condition which achieved complete silicon immobilization in the oxide scale in aqueous environments was the formation of ZrSiO4 via ZrSi2 coating oxidation in 1400 °C air. In addition, ZrSi2 coatings were beneficial in enhancing quenching heat transfer - the minimum film boiling temperature increased by 6-8% in the three different environmental conditions tested. During repeated thermal cycles (water quenching from 700 °C to 85 °C for 20 s) performed as a part of quench tests, no spallation and cracking was observed and the coating prevented oxidation of the underlying Zircaloy-4 substrate.

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-07-01

The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

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

Fuel cladding tube and fuel rod for BWR type reactor

International Nuclear Information System (INIS)

Urata, Megumu; Mitani, Shinji.

1995-01-01

A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

Modeling of Zircaloy cladding degradation under repository conditions

International Nuclear Information System (INIS)

Santanam, L.; Raghavan, S.; Chin, B.A.

1989-07-01

Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

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.

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.)

Mechanisms of fuel-cladding chemical interaction: US interpretation

International Nuclear Information System (INIS)

Adamson, M.G.

1977-01-01

Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

Mechanisms of fuel-cladding chemical interaction: US interpretation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States)

1977-04-01

Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

Influence of processing variables and alloy chemistry on the corrosion behavior of ZIRLO nuclear fuel cladding

International Nuclear Information System (INIS)

Comstock, R.J.; Sabol, G.P.; Schoenberger, G.

1996-01-01

Variations in the thermal heat treatments used during the fabrication of ZIRLO (Zr-1Nb-1Sn-0.1Fe) fuel clad tubing and in ZIRLO alloy chemistry were explored to develop a further understanding of the relationship between processing, microstructure, and cladding corrosion performance. Heat treatment variables included intermediate tube annealing temperatures as well as a beta-phase heat treatment during the latter stages of the tube reduction schedule. Chemistry variables included deviations in niobium and tin content from the nominal composition. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in both pure and lithiated water and high-temperature steam. Analytical electron microscopy demonstrated that the best out-reactor corrosion performance is obtained for microstructures containing a fine distribution of beta-niobium and Zr-Nb-Fe particles. Deviations from this microstructure, such as the presence of beta-zirconium phase, tend to degrade corrosion resistance. ZIRLO fuel cladding was irradiated in four commercial reactors. In all cases, the microstructure in the cladding included beta-niobium and Zr-Nb-Fe particles. ZIRLO fuel cladding processed with a late-stage beta heat treatment to further refine the second-phase particle size exhibited in-reactor corrosion behavior that was similar to reference ZIRLO cladding. Variations of the in-reactor corrosion behavior of ZIRLO were correlated to tin content, with higher oxide thickness observed in the ZIRLO cladding containing higher tin. The results of these studies indicate that optimum corrosion performance of ZIRLO is achieved by maintaining a uniform distribution of fine second-phase particles and controlled levels of tin

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

Energy Technology Data Exchange (ETDEWEB)

Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

2006-12-12

Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

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.

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

Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2008-01-01

Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

Fission product release in conditions of a spent fuel pool severe accident

International Nuclear Information System (INIS)

Ohai, Dumitru

2007-01-01

Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

International Nuclear Information System (INIS)

Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

1979-01-01

Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

Advanced in-situ characterisation of corrosion properties of LWR fuel cladding materials

International Nuclear Information System (INIS)

Arilahti, E.; Bojinov, M.; Beverskog, B.

1999-01-01

The trend towards higher fuel burnups imposes a demand for better corrosion and hydriding resistance of cladding materials. Development of new and improved cladding materials is a long process. There is a lack of fast and reliable in-situ techniques to investigate zirconium alloys in simulated or in-core LWR coolant conditions. This paper describes a Thin Layer Electrode (TLE) arrangement suitable for in-situ characterization of oxide films formed on fuel cladding materials. This arrangement enables us to carry out: Versatile Thin Layer Electrochemical measurements, including: (i) Thin Layer Electrochemical impedance Spectroscopic (TLEIS) measurements to characterize the oxidation kinetics and mechanisms of metals and the properties of their oxide films in aqueous environments. These measurements can also be performed in low conductivity electrolytes. (ii) Thin-Layer Wall-Jet (TLWJ) measurements, which give the possibility to detect soluble reaction products and to evaluate the influence of novel water chemistry additions on their release. Solid Contact measurements: (i) Contact Electric Resistance (CER) measurements to investigate the electronic properties of surface films on the basis of d.c. resistance measurements. (i) Contact Electric impedance (CEI) measurements to study the electronic properties of surface films using a.c. perturbation. All the above listed measurements can be performed using one single measurement device developed at VTT. This device can be conveniently inserted into an autoclave. Its geometry is currently being optimized in cooperation with the OECD Halden Reactor Project. In addition, the applicability of the device for in-core measurements has been investigated in a joint feasibility study performed by VTT and JRC Petten. Results of some autoclave studies of the effect of LiOH concentration on the stability of fuel cladding oxide films are presented in this paper. (author)

Ceramic Coatings for Clad (The C³ Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

Energy Technology Data Exchange (ETDEWEB)

Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

2017-02-14

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

Inspection system for Zircaloy clad fuel rods

International Nuclear Information System (INIS)

Yancey, M.E.; Porter, E.H.; Hansen, H.R.

1975-10-01

A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

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

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%.

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

International Nuclear Information System (INIS)

Roake, W.E.

1977-01-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

1977-04-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

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...

The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

International Nuclear Information System (INIS)

Smith, E.

1979-01-01

A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

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)

Review and evaluation of cladding attack of LMFBR fuel

International Nuclear Information System (INIS)

Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

1977-01-01

The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

Cladding properties under simulated fuel pin transients

International Nuclear Information System (INIS)

Hunter, C.W.; Johnson, G.D.

1975-01-01

A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

Fuel cladding mechanical interaction during power ramps

International Nuclear Information System (INIS)

Guerin, Y.

1985-01-01

Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

Fuel-cladding chemical interaction in mixed-oxide fuels

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, J.W.; Devary, J.L.

1978-10-01

The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

International Nuclear Information System (INIS)

McClelland, R.G.; O'Leary, P.M.

1992-01-01

Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an ∼0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current 'lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4

MAX Phase Modified SiC Composites for Ceramic-Metal Hybrid Cladding Tubes

International Nuclear Information System (INIS)

Jung, Yang-Il; Kim, Sun-Han; Park, Dong-Jun; Park, Jeong-Hwan; Park, Jeong-Yong; Kim, Hyun-Gil; Koo, Yang-Hyun

2015-01-01

A metal-ceramic hybrid cladding consists of an inner zirconium tube, and an outer SiC fiber-matrix SiC ceramic composite with surface coating as shown in Fig. 1 (left-hand side). The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. In addition, the outermost layer prevents the dissolution of SiC during normal operation. On the other hand, a ceramic-metal hybrid cladding consists of an outer zirconium tube, and an inner SiC ceramic composite as shown in Fig. 1 (right-hand side). The outer zirconium protects the fuel rod from a corrosion during reactor operation, as in the present fuel claddings. The inner SiC composite, additionally, is designed to resist the severe oxidation under a postulated accident condition of a high-temperature steam environment. Reaction-bonded SiC was fabricated by modifying the matrix as the MAX phase. The formation of Ti 3 SiC 2 was investigated depending on the compositions of the preform and melt. In most cases, TiSi 2 was the preferential phase because of its lowest melting point in the Ti-Si-C system. The evidence of Ti 3 SiC 2 was the connection with the pressurizing

Some recent trends in the use of zirconium alloys for nuclear service

International Nuclear Information System (INIS)

Balaramamoorthy, K.

1992-01-01

Without any exception nuclear power reactors particularly the water cooled ones, operating in the World use natural or slightly enriched uranium oxide fuel pellets with zirconium alloy cladding. While the zirconium alloys have proven to be successful in their designed usage, a desire for longer lifetimes of core components and increased duty cycle puts more demand on materials performance. This demand has led to more in depth studies of phenomena associated with zirconium alloy corrosion mechanism, fine tuning of the zirconium alloy composition, development of fabrication techniques and to the evaluation of newer zirconium alloys for critical applications. (author). 5 refs., 32 figs

Study on dynamic measurement of fuel pellet length during loading into cladding tube

International Nuclear Information System (INIS)

Zhang Kai

1993-09-01

Various methods are presented for measuring the pellet length in the cladding tube (zirconium tube) during the loading process of the preparation of single rod of nuclear fuel assembly. These methods are used in former Soviet Union, west European countries and China in the manufacturing of nuclear power plant element. Different methods of dynamic measurement by using mechanics, optics and electricity and their special features are analysed and discussed. The structure and measuring principle of a developed measuring device,and its measuring precision and system deviation are also introduced. Finally, the length of loaded pellets is checked with analog pellets. The results are as expected and show that the method and principle used in the measuring device are feasible. It is an ideal and advanced method for the pellet loading of single cladding tube. The principle mentioned above can also be used in other industries

Experimental assessment of fuel-cladding interactions

Energy Technology Data Exchange (ETDEWEB)

Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-29

A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

Production of nuclear grade zirconium: A review

Energy Technology Data Exchange (ETDEWEB)

Xu, L., E-mail: L.Xu-2@tudelft.nl [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands); Xiao, Y. [Department of Metallurgical Engineering, Anhui University of Technology, Ma' anshan 243002 (China); Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Sandwijk, A. van [Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Xu, Q. [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Yang, Y. [Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands)

2015-11-15

Zirconium is an ideal material for nuclear reactors due to its low absorption cross-section for thermal neutrons, whereas the typically contained hafnium with strong neutron-absorption is very harmful for zirconium as a fuel cladding material. This paper provides an overview of the processes for nuclear grade zirconium production with emphasis on the methods of Zr–Hf separation. The separation processes are roughly classified into hydro- and pyrometallurgical routes. The known pyrometallurgical Zr–Hf separation methods are discussed based on the following reaction features: redox characteristics, volatility, electrochemical properties and molten salt–metal equilibrium. In the present paper, the available Zr–Hf separation technologies are compared. The advantages and disadvantages as well as future directions of research and development for nuclear grade zirconium production are discussed.

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.)

Development in the manufacture of fuel assembly components at Nuclear Fuel Complex

International Nuclear Information System (INIS)

Saibaba, N.

2012-01-01

The integrity of the fuel bundle and pellet-clad mechanical and chemical interaction (PCMCI) is the major limiting factor in achieving high burn up in thermal as well as fast reactors. Zircaloy based fuel bundle used for Indian pressurized heavy water reactor consists of number of components such as fuel clad tube, end cap bearing pad and spacer pad. These tubular, bar and sheet components are manufactured at Nuclear Fuel Complex using a series of thermomechanical processes involving hot and cold working with intermediate heat treatment. This paper is aimed at bringing out recent advances in NFC in the manufacture of fuel assembly components. Zircaloy based double clad tube adopting co-extrusion route followed by cold pilgering was successfully produced for its potential usage for high burnup in advance thermal reactors such as Advanced Heavy Water Reactors, This paper also includes process modifications carried out in the manufacture of clad tube and end cap components based on in-depth metallurgical studies. A radial forging process was established for primary breakdown of arc melted ingot which allows for better soundness and homogeneous microstructure. Manufacturing route of bar components for end caps was suitably modified by adopting only barrel straightening to minimize the residual stress and thereby increasing the recovery appreciably. NFC also supplies clad tube for fast breeder reactors where limiting factor for burn up are void swelling and fuel-clad interaction. In view of this, advance claddings such as P/M based 9Cr - Oxide Dispersion strengthened (ODS) steel clad and Zirconium lined T91 (9Cr-1 Mo) steel double clad have been successfully produced. Zirconium lined T91 (9Cr-1 Mo) double clad tubes required was successfully produced by adopting the method of co-pilgering, as a candidate material for clad tubes of Fast Breeder Reactors. (author)

Investigating mechanical behavior and radiation resistant of fuel rods clad in nuclear power plant

International Nuclear Information System (INIS)

Sedgh Kerdar, A.

1999-01-01

interstitials in metal lattice under irradiation causes increased strength and hardness but decreases ductility in metals.The increase in strength and hardness depends on obstacles that prevent the motion of dislocations. The clustering of point defects are responsible for these changes. Irradiation also induces instabilities in phases due to enhancement of diffusion, solute segregation, precipitate formation, order- disorder transformation and resolution of small precipitates. From the microscopic point of view accumulation of vacancies accompanied by formation of He and H 2 gases under irradiation cause an increase in volume which results in swelling and eventually ends up with embrittlement of metals. This subject was described in chapter three Zirconium and its alloys are the best structural materials for fuel cladding of BWR and PWR reactors core. The working condition in the core of nuclear reactor are very serve, respect temperature and radiation dose. It should be realized that, if fuel cladding receive damage and get cracked, the first cooling cycle and the maine equipment will be contaminated with active materials which cause additional environmental problems. Furthermore, replacement of fuel rods are very costly. Therefore, for increasing life time of fuel cladding and minimizing damage, the effect of radiation and heat on Zirconium and its alloys must be investigated. This subject was described in chapter four.The mechanical behavior and radiation resistant of fuel cladding in PWR reactor (specifically WWER ) have been investigated which is described in chapter five. Result, discussion and final conclusion are summarized in last chapter and also several points for improvement have been offered

Pellet-clad interaction in water reactor fuels

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

Pellet-clad interaction in water reactor fuels

International Nuclear Information System (INIS)

2004-01-01

The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

UK experience on fuel and cladding interaction in oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

1977-04-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

UK experience on fuel and cladding interaction in oxide fuels

International Nuclear Information System (INIS)

Batey, W.; Findlay, J.R.

1977-01-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

Fuel compliance model for pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Shah, V.N.; Carlson, E.R.

1985-01-01

This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

Phase evolution and dendrite growth in laser cladding of aluminium on zirconium

International Nuclear Information System (INIS)

Yue, T.M.; Xie, H.; Lin, X.; Yang, H.O.

2011-01-01

Research highlights: → Laser cladding of Al on pure Zr. → A series of phase evolutions occurred across the laser-clad coating. → Epitaxial crystal growth, backward dendrite growth and two-phase eutectic dendritic growth. → Phase and microstructure evolution is discussed. - Abstract: Aluminium was laser clad on a pure zirconium substrate using the blown powder method. The microstructure across the laser-clad coating was studied. Starting from the bottom to the top surface of the coating, a series of phase evolutions had occurred: (Zr) → (Zr) + AlZr 2 + AlZr 3 → Al 4 Zr 5 + Al 3 Zr 2 → Al 3 Zr 2 + AlZr 2 → Al 2 Zr → Al 2 Zr + Al 3 Zr. This resulted in an epitaxial columnar crystal growth at the re-melt substrate boundary, a band of backward growth Al 3 Zr 2 dendrites towards the lower half of the coating, and a two-phase eutectic dendritic growth of Al 2 Zr + Al 3 Zr towards the top of the coating. The evolution of the various phases and microstructures is discussed in conjunction with the Al-Zr phase diagram, the criteria for planar interface instability, and the theory of eutectic growth under rapid solidification conditions (the TMK model).

Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

2016-12-15

Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

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

Effects of cold worked and fully annealed claddings on fuel failure behaviour

International Nuclear Information System (INIS)

Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi

1979-12-01

Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident (RIA) condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. (author)

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

The ballooning of fuel cladding tubes: theory and experiment

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1988-01-01

Under some conditions, fuel clad ballooning can result in considerable strain before rupture. If ballooning were to occur during a loss-of-coolant accident (LOCA), the resulting substantial blockage of the sub-channel would restrict emergency core cooling. However, circumferential temperature gradients that would occur during a LOCA may significantly limit the average strain at failure. Understandably, the factors that control ballooning and rupture of fuel clad are required for the analysis of a LOCA. Considerable international effort has been spent on studying the deformation of Zircaloy fuel cladding under conditions that would occur during a LOCA. This effort has established a reasonable understanding of the factors that control the ballooning, failure time, and average failure strain of fuel cladding. In this paper, both the experimental and theoretical studies of the fuel clad ballooning are reviewed. (author)

State-of-the-technology review of fuel-cladding interaction

International Nuclear Information System (INIS)

Bailey, W.J.; Wilson, C.L.; MacGowan, L.J.; Pankaskie, P.J.

1977-12-01

A literature survey and a summarization of postulated fuel-cladding-interaction mechanisms and associated supportive data are reported. The results of that activity are described in the report and include comments on experience with power-ramped fuel, fuel-cladding mechanical interaction, stress-corrosion cracking and fission-product embrittlement, potential remedial actions, fuel-cladding-interaction mechanistic considerations, other ongoing programs, and related patents of interest. An assessment of the candidate fuel concepts to be evaluated as part of this program is provided

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

Tube in zirconium base alloy for nuclear fuel assembly and manufacturing process of such a tube

International Nuclear Information System (INIS)

Mardon, J.P.; Senevat, J.; Charquet, D.

1996-01-01

This patent concerns the description and manufacturing guidelines of a zirconium alloy tube for fuel cladding or fuel assembly guiding. The alloy contains (in weight) 0.4 to 0.6% of tin, 0.5 to 0.8% of iron, 0.35 to 0.50% of vanadium and 0.1 to 0.18% of oxygen. The carbon and silicon tenors range from 100 to 180 ppm and from 80 to 120 ppm, respectively. The alloy contains only zirconium, plus inevitable impurities, and is completely recrystallized. Corrosion resistance tests were performed on tubes made of this alloy and compared to corrosion tests performed on zircaloy 4 tubes. These tests show a better corrosion resistance and a lower corrosion kinetics for the new alloy, even in presence of lithium and iodine, and a lower hydridation rate. The mechanical resistance of this alloy is slightly lower than the one of zircaloy 4 but becomes equivalent or slightly better after two irradiation cycles. The ductility remains always equal or better than for zircaloy 4. (J.S.)

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

Computer analysis of elongation of the WWER fuel rod claddings

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2008-01-01

In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

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.)

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

Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

International Nuclear Information System (INIS)

Minato, K.; Hering, W.; Hagen, S.

1991-07-01

Out-of-pile bundle experiments (zircaloy 4) are performed in the CORA facility to investigate the behavior of PWR fuel elements during severe fuel damage (SFD) accidents. Within the international cooperation the most significant phenomena such as cladding deformation, oxidation (especially the zirconium/steam reaction), melt formation, melt release, and relocation which were found in all tests have been analyzed. (orig./MM) [de

Cladding embrittlement during postulated loss-of-coolant accidents.

Energy Technology Data Exchange (ETDEWEB)

Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

2008-07-31

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

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

Performance of refractory alloy-clad fuel pins

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

1984-12-01

This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

1994-10-01

This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

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 �

Development Status of Accident Tolerant Fuel Cladding for LWRs

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

Laves intermetallics in stainless steel-zirconium alloys

International Nuclear Information System (INIS)

Abraham, D.P.; McDeavitt, S.M.; Richardson, J.W. Jr.

1997-01-01

Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni) 2+x , have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni) 23 Zr 6 during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy

Nuclear fuel element, and method of producing same

International Nuclear Information System (INIS)

Armijo, J.S.; Esch, E.L.

1986-01-01

This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

Fuel clad chemical interactions in fast reactor MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Viswanathan, R., E-mail: rvis@igcar.gov.in

2014-01-15

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ⋅ [B/(at.% fission)] ⋅ (T/K-705) ⋅ [(O/M)_i-1.935]} + 20.5) for (O/M){sub i} ⩽ 1.98. A new model is proposed for (O/M){sub i} ⩾ 1.98: d/μm = [B/(at.% fission)] ⋅ (T/K-800){sup 0.5} ⋅ [(O/M){sub i}-1.94] ⋅ [P/(W cm{sup −1})]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M){sub i} is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

Secondary hydriding of defected zircaloy-clad fuel rods

International Nuclear Information System (INIS)

Olander, D.R.; Vaknin, S.

1993-01-01

The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

Stainless steel clad for light water reactor fuels. Final report

International Nuclear Information System (INIS)

Rivera, J.E.; Meyer, J.E.

1980-07-01

Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

Fuel cladding behavior under rapid loading conditions

Science.gov (United States)

Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

2016-02-01

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

Mechanical modelling of transient- to- failure SFR fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Feria, F.; Herranz, L. E.

2014-07-01

The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

Oxidation Behavior of FeCrAl -coated Zirconium Cladding prepared by Laser Coating

Energy Technology Data Exchange (ETDEWEB)

Kim, Il-Hyun; Kim, Hyun-Gil; Choi, Byung-Kwan; Park, Jeong-Yong; Koo, Yang-Hyun; Kim, Jin-Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

From the recent research trends, the ATF cladding concepts for enhanced accident tolerance are divided as follows: Mo-Zr cladding to increase the high temperature strength, cladding coating to increase the high temperature oxidation resistance, FeCrAl alloy and SiC/SiCf material to increase the oxidation resistance and strength at high temperature. To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. A laser coating method supplied with FeCrAl powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a FeCrAl-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Cladding failure margins for metallic fuel in the integral fast reactor

International Nuclear Information System (INIS)

Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

1987-01-01

The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

In-reactor fuel cladding external corrosion measurement process and results

International Nuclear Information System (INIS)

Thomazet, J.; Musante, Y.; Pigelet, J.

1999-01-01

Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

Advanced ceramic cladding for water reactor fuel

International Nuclear Information System (INIS)

Feinroth, H.

2000-01-01

Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

Prevention of nuclear fuel cladding materials corrosion

International Nuclear Information System (INIS)

Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

1983-01-01

The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

Energy Technology Data Exchange (ETDEWEB)

Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

1997-02-01

Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

Examination of Zircaloy-clad spent fuel after extended pool storage

International Nuclear Information System (INIS)

Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

1981-09-01

This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed

Management of cladding hulls and fuel hardware

International Nuclear Information System (INIS)

1985-01-01

The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

Clad Treatment in KARMA Code and Library

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

2016-05-15

Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

Fuel-clad heat transfer coefficient of a defected fuel rod

International Nuclear Information System (INIS)

Bruet, M.; Stora, J.P.

1976-01-01

A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

Delayed hydride cracking of Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

2007-01-01

Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-01

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

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.)

The prediction problems of VVER fuel element cladding failure theory

International Nuclear Information System (INIS)

Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

2016-01-01

Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

Fuel cladding mechanical properties for transient analysis

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

1976-01-01

Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

Characteristics of hydride precipitation and reorientation in spent-fuel cladding

International Nuclear Information System (INIS)

Chung, H. M.; Strain, R. V.; Billone, M. C.

2000-01-01

The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

Corrosion effect of fast reactor fuel claddings on their mechanical properties

International Nuclear Information System (INIS)

Davydov, E.F.; Krykov, F.N.; Shamardin, V.K.

1985-01-01

Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature

General considerations on the oxide fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Pascard, R.

1977-01-01

Since the very first experimental irradiations in thermal reactors, performed in view of the future Rapsodie fuel general study, corrosion cladding anomalies were observed. After 10 years of Rapsodie and more than two years of Phenix, performance brought definite confirmation of the chemical reactions between the irradiated fuel and cladding. That is the reason for which the fuel designers express an urgent need for determining the corrosion rates. Semi-empirical laws and mechanisms describing corrosion processes are proposed. Erratic conditions for appearance of the oxide-cladding corrosion are stressed upon. Obviously such a problem can be fully appreciated only by a statistical approach based on a large number of observations on the true LMFBR fuel pins

Multidimensional simulations of hydrides during fuel rod lifecycle

International Nuclear Information System (INIS)

Stafford, D.S.

2015-01-01

In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim. - Highlights: • We extend BISON fuel performance code to simulate lifecycle of fuel rods. • We model hydrogen evolution in cladding from reactor through dry storage. • We validate 1D simulations of hydrogen evolution against experiments. • We show results of 2D axisymmetric simulations predicting hydride formation. • We show how our model predicts formation of a hydride rim in the cladding.

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

Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

International Nuclear Information System (INIS)

Chen Daolong

1986-01-01

A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels

International Nuclear Information System (INIS)

Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R.

1996-01-01

An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air

Modelling of pellet-cladding interaction for PWRs reactors fuel rods

International Nuclear Information System (INIS)

Esteves, A.M.

1991-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

Probabilistic assessment of spent-fuel cladding breach

International Nuclear Information System (INIS)

Foadian, H.; Rashid, Y.R.; Seager, K.D.

1991-01-01

A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B ampersand W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

Probabilistic assessment of spent-fuel cladding breach

International Nuclear Information System (INIS)

Foadian, H.; Rashid, Y.R.; Seager, K.D.

1992-01-01

In this paper a methodology for determining the probability of spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B and W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

Metallic fuel development

International Nuclear Information System (INIS)

Walters, L.C.

1987-01-01

Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Modelling the gas transport and chemical processes related to clad oxidation and hydriding

Energy Technology Data Exchange (ETDEWEB)

Montgomery, R O; Rashid, Y R [ANATECH Research Corp., San Diego, CA (United States)

1997-08-01

Models are developed for the gas transport and chemical processes associated with the ingress of steam into a LWR fuel rod through a small defect. These models are used to determine the cladding regions in a defective fuel rod which are susceptible to massive hydriding and the creation of sunburst hydrides. The brittle nature of zirconium hydrides (ZrH{sub 2}) in these susceptible regions produces weak spots in the cladding which can act as initiation sites for cladding cracks under certain cladding stress conditions caused by fuel cladding mechanical interaction. The modeling of the axial gas transport is based on gaseous bimolar diffusion coupled with convective mass transport using the mass continuity equation. Hydrogen production is considered from steam reaction with cladding inner surface, fission products and internal components. Eventually, the production of hydrogen and its diffusion along the length results in high hydrogen concentration in locations remote from the primary defect. Under these conditions, the hydrogen can attack the cladding inner surface and breakdown the protective ZrO{sub 2} layer locally, initiating massive localized hydriding leading to sunburst hydride. The developed hydrogen evolution model is combined with a general purpose fuel behavior program to integrate the effects of power and burnup into the hydriding kinetics. Only in this manner can the behavior of a defected fuel rod be modeled to determine the conditions the result in fuel rod degradation. (author). 14 refs, 6 figs.

Fabrication of oxide dispersion strengthened ferritic clad fuel pins

International Nuclear Information System (INIS)

Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

1991-01-01

A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

Electrochemical Studies on Important Elements for Zirconium Recovery Form Irradiated Zircaloy-4 Cladding

International Nuclear Information System (INIS)

Park, J.; Sohn, S.; Hwang, I.S.

2015-01-01

Since Zircaloy cladding accounts for about 16 wt. % of used nuclear fuel assembly, decontamination process is required to reduce the final waste volume from spent nuclear fuel. To develop Zircaloy-4 electrorefining process as an irradiated Zircaloy cladding decontamination process, electrochemical studies on Sn, Cr, Fe and Co which are major or important elements in the irradiated cladding were conducted based on cyclic voltammetry in LiCl-KCl at 500 deg. C. Cyclic voltammetry for Sn, Fe, Cr and Co elements that should be eliminated was conducted and revealed that redox reactions of these ions are much simpler than Zr and more reductive than Zr. The reliability of cyclic voltammetry was verified by comparing diffusion coefficients and formal reduction potentials of these ions obtained in this study to previous studies. (authors)

Creep collapse of TAPS fuel cladding

International Nuclear Information System (INIS)

Chaudhry, S.M.; Anand, A.K.

1975-01-01

Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

Gap conductance in Zircaloy-clad LWR fuel rods

International Nuclear Information System (INIS)

Ainscough, J.B.

1982-04-01

This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

Influence of the fuel operational parameters on the aluminium cladding quality of discharged fuel

Energy Technology Data Exchange (ETDEWEB)

Chwaszczewski, S.; Czajkowski, W.; Borek-Kruszewska, E. [Institute of Atomic Energy, Otwock Swierk (POLAND)

2002-07-01

In the last two years, the new MR6 type fuel containing 1550 g of U with 36% enrichment has been loaded into MARIA reactor core. Its aluminium cladding thickness is 0,6 mm and typical burnup -about 4080 MWh (as compared to 2880 MWh for the 80% enriched fuel used). However, increased fission product release from these assemblies was observed near the end of its operational time. The results presented earlier [1] show that the corrosion behaviour of aluminium cladding depends on the conditions of fuel operation in the reactor. The corrosion process in the aluminum of fuel cladding proceeds faster then in the aluminum of constructional elements. This tendency was also observed in MR-6/80% and in WWR- SM fuel assemblies. Therefore the visual tests of discharged MR-6/36% fuel elements were performed. Some change of appearance of aluminum cladding was observed, especially in the regions with large energy generation i.e. in the centre of reactor core and in the strong horizontal gradient of neutron flux. In the present paper, the results of visual investigation of discharged fuel assemblies are presented. The results of the investigation are correlated with the operational parameters. (author)

Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

International Nuclear Information System (INIS)

Mazeres, Benoit

2013-01-01

One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

Energy Technology Data Exchange (ETDEWEB)

Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

1999-04-01

To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2009-12-01

A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

First results on the effect of fuel-cladding eccentricity

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2009-01-01

In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. (Authors)

Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1979-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

1979-01-01

This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

Energy Technology Data Exchange (ETDEWEB)

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

Fuel Performance Calculations for FeCrAl Cladding in BWRs

Energy Technology Data Exchange (ETDEWEB)

George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

Study of pellet clad interaction defects in Dresden-3 fuel rods

International Nuclear Information System (INIS)

Pasupathi, V.; Perrin, J.S.

1979-01-01

During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

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.)

Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

International Nuclear Information System (INIS)

2000-10-01

The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

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

International Nuclear Information System (INIS)

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

1992-10-01

U-3Zr and U-20.5Pu-3Zr were injection cast into Zr tubes, or sheaths, rather than into quartz molds and clad in 316SS. These elements and standard-cast U-l0Zr and U-IgPu-l0Zr elements were irradiated in EBR-II to 2 at.% and removed for interim examination. Measurements of axial growth at indicate that the Zr-sheathed elements exhibited significantly less axial elongation than the standard-cast elements (1.3 to 1.8% versus 4.9 to 8.1%). Fuel material extruded through the ends of the Zr sheaths. allowing the low-Zr fuel to contact the cladding in some cases. Transverse metallographic sections reveal cracks in the Zr sheath through which fuel extruded and contacted cladding. The sheath is not a sufficient barrier between fuel and cladding to reduce FCCI. and any adverse effects due to increased FCCI will be evident as the elements attain higher burnup

Performance of HT9 clad metallic fuel at high temperature

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

1992-01-01

Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

International Nuclear Information System (INIS)

Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

2010-01-01

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

Influence of pellet-clad-gap-size on LWR fuel rod performance

International Nuclear Information System (INIS)

Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

1979-01-01

The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

In-reactor performance of methods to control fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

1979-01-01

Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

Mechanical and temperature contact in fuel rod cladding

International Nuclear Information System (INIS)

Fredriksson, B.E.; Rydholm, S.G.

1977-01-01

The paper presents results for the effect of different types of slip rules on the contact stress distribution. It is shown that the contact shear stress is smaller for the hardening model than for the ideal model. It is also shown that a crack in the fuel increases the contact stresses and that at temperature decrease high tensile stresses arise after eventual welding. It is also shown how particles between fuel and cladding influence the stresses. Also here the effect of eventual welding is studied. The present method is well suited to study cracks and crack propagation. The surfaces of the existing cracks are defined as contact surfaces and the crack extension work is calculated by releasing the nodes at the crack tip. As the crack surfaces are defined as contact surfaces eventual crack closure is automatically taken into account. Crack extension work is calculated for existing cracks in the cladding. It is shown that cracks in the fuel and particles between fuel and cladding will increase the crack extension work

Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins

International Nuclear Information System (INIS)

Lawrence, L.A.

1986-10-01

A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding

Radiochemical neutron activation analysis of zirconium and zirconium-niobium alloys

International Nuclear Information System (INIS)

Tashimova, F.A.; Sadikov, I.I.; Salimov, M.

2004-01-01

Full text: Zirconium and zirconium-niobium alloys are used on nuclear technology, as fuel cladding of nuclear reactors. Their nuclear-physical, mechanical and thermophysical properties are influenced them matrix and impurity composition, therefore determination of matrix and impurity content of these materials is a very important task. Neutron activation analysis is one from multielemental and high sensible techniques that are widely applied in analysis of high purity materials. Investigation of nuclear-physical characteristics of zirconium has shown that instrumental variant NAA is unusable for analysis due to high radioactivity of a matrix. Therefore it is necessary carrying out radiochemical separation of impurity radionuclides from matrix. Study of the literature datum have shown, that zirconium and niobium are very well extracted from muriatic solution with 5% tributyl phosphineoxide (TBPO) solution in toluene and 0,75 M solution of di-2-ethyl hexyl phosphoric acid (HDEHP) in cyclohexanone. Investigation of these elements extraction in these systems has shown that more effective and selective separation of matrix radionuclides is achieved in HDEHP-3M HCI system. This system is also extracted and hafnium, witch is an accompanying element of zirconium and its high content prevented determination of other impurity elements in sample. Therefore we used extraction system HDEHP-3M HCl for analysis of zirconium and zirconium-niobium alloys in chromatographic variant. By measurement of distribution profile of a matrix and of elution curve of determined elements is established, that for effective separation of impurity and matrix radionuclides there is enough chromatographic column with diameter 1 cm and height of a sorbent layer 7 cm, thus volume of elute, necessary for complete elution of determinate elements is 35-40 ml. On the basis of the carried out researches the technique of radiochemical NAA of high purity zirconium and zirconium-niobium alloy, which allows to

Design of an integrated system to recycle Zircaloy cladding using a hydride–milling–dehydride process

Energy Technology Data Exchange (ETDEWEB)

Kelley, Randy, E-mail: rkelley@pitt.edu [Mechanical Engineering Department, 236 Engineering and Science Building, University of Pittsburgh – Johnstown, Johnstown, PA 15904 (United States); McDeavitt, Sean [Texas A and M University, Department of Nuclear Engineering, 327 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

2013-10-15

Highlights: • Dehydriding zirconium hydride was studied at relatively low temperatures (<800 °C). • High vacuum pressures decrease dehydriding temperatures. • Specialized equipment was designed, built and demonstrated to process zirconium. • The process hydrided–milled–dehydrided zirconium metal to a fine metal powder. • Two powder samples were analyzed and proved the operation of the machine. -- Abstract: A hydride–dehydride process was evaluated to recover a portion of spent nuclear fuel cladding; a zirconium alloy (Zircaloy), as a metal powder that may be used for advanced nuclear fuel applications. The investigation was part of a broader study that sought to determine the viability of recovering components of used nuclear fuel to for a metal matrix cermet for transuranic burning. The zirconium powder process begins with the conversion of Zircaloy cladding hulls into a brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. In support of this, a specialized piece of equipment was designed to demonstrate the entire zirconium conversion process to transform Zircaloy tubes into metal powder without intermediate handling. This was accomplished by building a milling system that rotates inside of controlled atmosphere chamber with an internal heater. The hydriding process was accomplished using an argon–5% hydrogen atmosphere at 500 °C. The process variables for the dehydriding process were determined using a thermogavimetric analysis (TGA) method. It was determined that a rough vacuum (∼0.001 bar) and 800 °C were sufficient to decompose the zirconium hydride. Zirconium metal powder was created using different milling times: 45 min (coarse powder) and 12 h (fine powder). X-ray diffraction (XRD) analysis indicated that the process produced a zirconium metal. Additionally, visual observations of the samples silvery

Long term wet spent nuclear fuel storage

International Nuclear Information System (INIS)

1987-04-01

The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

International Nuclear Information System (INIS)

Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

2010-01-01

Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

In-reactor cladding breach of EBR-II driver-fuel elements

International Nuclear Information System (INIS)

Seidel, B.R.; Einziger, R.E.

1977-01-01

Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

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

FRACAS: a subcode for the analysis of fuel pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Bohn, M.P.

1977-04-01

This report describes FRACAS (Fuel Rod and Cladding Analysis Subcode), a computer code which performs the mechanical analysis in the FRAP fuel rod codes. At each loadstep, FRACAS obtains a complete elastic-plastic-creep solution for the stresses, strains, and displacements in the fuel rod cladding. The cladding is modeled as a thin cylindrical shell with prescribed temperature, pressures, and radial displacement of the inside surface. The displacement of the fuel pellets is assumed to be due to thermal gradients only. Three different regimes of pellet-cladding mechanical interaction are considered: (a) open gap, (b) closed gap, and (c) trapped stack. Both transient and steady state creep calculations are performed. The capabilities of the code are illustrated by an example problem, and comparisons are made with data obtained from two experimental fuel rods

Chemical interaction of fuel and cladding tubes

International Nuclear Information System (INIS)

Kirihara, Tomoo; Yamawaki, Michio; Obata, Naomi; Handa, Muneo.

1983-01-01

It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. (Kako, I.)

Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

International Nuclear Information System (INIS)

Wilson, W.; Comstock, R.J.

1999-01-01

Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

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.

Irradiation experience with HT9-clad metallic fuel

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

1991-01-01

The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-01-01

The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

Facility for in-reactor creep testing of fuel cladding

International Nuclear Information System (INIS)

Kohn, E.; Wright, M.G.

1976-11-01

A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

Microbial biofilm growth on irradiated, spent nuclear fuel cladding

International Nuclear Information System (INIS)

Bruhn, D.F.; Frank, S.M.; Roberto, F.F.; Pinhero, P.J.; Johnson, S.G.

2009-01-01

A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

High performance fuel technology development : Development of high performance cladding materials

International Nuclear Information System (INIS)

Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

2012-04-01

The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

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

Siemens advance PWR fuel assemblies (HTP) and cladding

International Nuclear Information System (INIS)

Stout, R. B.; Woods, K. N.

1997-01-01

This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available

Temperature distribution determination of JPSR power reactor fuel element and cladding

International Nuclear Information System (INIS)

Sudarmono

1996-01-01

In order to utilize of fuel rod efficiency, a concept of JAERI passive Safety Reactor (JPSR) has been developed in Japan Atomic Energy Research Institute. In the JPSR design, UO 2 . are adopted as a fuel rod. The temperature distribution in the fuel rod and cladding in the hottest channel is a potential limiting design constraint of the JPSR. In the present determination, temperature distribution of the fuel rod and cladding for JPSR were PET:formed using COBRA-IV-I to evaluate the safety margin of the present JPSR design. In this method, the whole core was represented by the 1/4 sector and divided into 50 subchannels and 40 axial nodes. The temperature become maximum at the elevation of 1.922 and 2.196 m in the typical cell under operating condition. The maximum temperature in the center of the fuel rod surface of the fuel rod and cladding were 1620,4 o C, 722,8 o C, and 348,6 o C. The maximum results of temperature in the center of the fuel rod and cladding; were 2015,28 o C and 550 o C which were observed at 3.1 second in the typical cell

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

Technical committee meeting on fuel and cladding interaction. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-04-01

Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases.

Technical committee meeting on fuel and cladding interaction. Summary report

International Nuclear Information System (INIS)

1977-04-01

Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

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)

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Acceptance criteria for interim dry storage of aluminum-clad fuels

International Nuclear Information System (INIS)

Sindelar, R.L.; Peacock, H.B. Jr.; Iyer, N.C.; Louthan, M.R. Jr.

1994-01-01

Direct repository disposal of foreign and domestic research reactor fuels owned by the United States Department of Energy is an alternative to reprocessing (together with vitrification of the high level waste and storage in an engineered barrier) for ultimate disposition. Neither the storage systems nor the requirements and specifications for acceptable forms for direct repository disposal have been developed; therefore, an interim storage strategy is needed to safely store these fuels. Dry storage (within identified limits) of the fuels received from wet-basin storage would avoid excessive degradation to assure post-storage handleability, a full range of ultimate disposal options, criticality safety, and provide for maintaining confinement by the fuel/clad system. Dry storage requirements and technologies for US commercial fuels, specifically zircaloy-clad fuels under inert cover gas, are well established. Dry storage requirements and technologies for a system with a design life of 40 years for dry storage of aluminum-clad foreign and domestic research reactor fuels are being developed by various groups within programs sponsored by the DOE

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Performance testing of refractory alloy-clad fuel elements for space reactors

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

1985-01-01

Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

Highly corrosion resistant zirconium based alloy for reactor structural material

International Nuclear Information System (INIS)

Ito, Yoichi.

1996-01-01

The alloy of the present invention is a zirconium based alloy comprising tin (Sn), chromium (Cr), nickel (Ni) and iron (Fe) in zirconium (Zr). The amount of silicon (Si) as an impurity is not more than 60ppm. It is preferred that Sn is from 0.9 to 1.5wt%, that of Cr is from 0.05 to 0.15wt%, and (Fe + Ni) is from 0.17 to 0.5wt%. If not less than 0.12wt% of Fe is added, resistance against nodular corrosion is improved. The upper limit of Fe is preferably 0.40wt% from a view point of uniform suppression for the corrosion. The nodular corrosion can be suppressed by reducing the amount of Si-rich deposition product in the zirconium based alloy. Accordingly, a highly corrosion resistant zirconium based alloy improved for the corrosion resistance of zircaloy-2 and usable for a fuel cladding tube of a BWR type reactor can be obtained. (I.N.)

Circumferential nonuniformity of cladding radiation swelling of fast reactor peripheral fuel elements

International Nuclear Information System (INIS)

Reutov, V.F.; Farkhutdinov, K.G.

1977-01-01

The results are presented of the investigation into the perimeter radiation swelling of Kh18N10T stainless steel cladding in different cross sections of a peripheral fuel element of the BR-5 reactor. The fluence on the cladding is 1.8-2.9 x 10 22 fast neutr/cm 2 , the operating temperatures in different parts of the fuel element being 430 deg to 585 deg C. There has been observed circumferential non-uniformity of the distribution, concentration, and of the total volume of radiation cavities, which is due to temperature non-uniformity along the cladding perimeter. It is shown that such non-uniformity of radiation swelling of the cladding material may result in bending of the peripheral fuel element with regard to the fuel assembly sheath walls

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

International Nuclear Information System (INIS)

El-Genk, M.S.; Croucher, D.W.

1979-01-01

The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

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)

Early implementation of SiC cladding fuel performance models in BISON

Energy Technology Data Exchange (ETDEWEB)

Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-18

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

1998-01-01

Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

Fuel-to-cladding heat transfer coefficient into reactor fuel element

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

Study on the improvement of nuclear fuel cladding reliability

International Nuclear Information System (INIS)

Rheem, Karp Soon; Han, Jung Ho; Jeong, Yong Hwan; Lee, Deok Hyun

1987-12-01

In order to improve the nuclear fuel cladding reliability for high burn-up fuels, the corrosion resistance of laser beam surface treated and β-quenched zircaloys and the mechanical characteristics including fatigue, burst, and out-of-pile PCMI characteristics of heat treated zircaloys were investigated. In addition, the inadiation characteristics of Ko-Ri reactor fuel claddings was examined. It was found that the wasteside corrosion resistance of commercial zircaloys was improved remarkably by laser beam surface treatment. The out-of-pile transient cladding failures were investigated in terms of hoop stress versus time-to-failures by means of mandrel loading units at 25 deg C and 325 deg C. Fatigue characteristics of the β-quenched and as-received zircaloy cladding were investigated by using an internal oil pressurization method which can simulate the load-following operation cycle. The results were in good agreement with the existing data obtained by conventional methods for commercial zircaloys. Burst tests were performed with commercial and the β-quenched zircaloys in high pressure argon gas atmosphere as a function of burst temperature. The burst stress decreased linearly in the α phase region up to 600 deg C and hereafter the decrement of the burst stress decreased gradually with temperature in the β-phase region. For the first time, the burst characteristic of the irradiated zircaloy-4 cladding tubes released from Ko-Ri nuclear power unit 1 was investigated, and attempts were made to trace the cause of cladding failures by examining the failed structure and fret marks by debris. (Author)

Statistical mechanical analysis of LMFBR fuel cladding tubes

International Nuclear Information System (INIS)

Poncelet, J.-P.; Pay, A.

1977-01-01

The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

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

Fuel System Compatibility Issues for Prometheus-1

International Nuclear Information System (INIS)

DC-- Noe; KB Gibbard; MH Krohn

2006-01-01

Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO 2 as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO 2 -based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

Experimental Setup with Transient Behavior of Fuel Cladding of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

Corrosion of research reactor aluminium clad spent fuel in water. Additional information

International Nuclear Information System (INIS)

2009-12-01

A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

Cladding failure margins for metallic fuel in the integral fast reactor

International Nuclear Information System (INIS)

Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

1987-01-01

The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

International Nuclear Information System (INIS)

Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

1979-01-01

Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

Pie technique of LWR fuel cladding fracture toughness test

International Nuclear Information System (INIS)

Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

2006-01-01

Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

International Nuclear Information System (INIS)

Park, Moon Ghu; Lee, Sang Hee

1997-01-01

As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

Science.gov (United States)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

A thermodynamic model for the attack behaviour in stainless steel clad oxide fuel pins

International Nuclear Information System (INIS)

Goetzmann, O.

1979-01-01

So far, post irradiation examination of burnt fuel pins has not revealed a clear cut picture of the cladding attack situation. For seemingly same conditions sometimes attack occurs, sometimes not. This model tries to depict the reaction possibilities along the inner cladding wall on the basis of thermodynamic facts in the fuel pin. It shows how the thermodynamic driving force for attack changes along the fuel column, and with different initial and operational conditions. Two criteria for attack are postulated: attack as a result of the direct reaction of reactive elements with cladding components; and attack as a result of the action of a special agent (CsOH). In defining a reaction potenial the oxygen potential, the temperature conditions (cladding temperature and fuel surface temperature), and the fission products are involved. For the determination of the oxygen potential at the cladding, three models for the redistribution of oxygen across the fuel/clad gap are offered. The effect of various parameters, like rod power, gap conductance, oxygen potential, inner wall temperature, on the thermodynamic potential for attack is analysed. (Auth.)

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)

Electroless deposition process for zirconium and zirconium alloys

Science.gov (United States)

Donaghy, Robert E.; Sherman, Anna H.

1981-01-01

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.

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)

Out-of-pile experiments of fuel-cladding chemical interaction, (2)

International Nuclear Information System (INIS)

Konashi, Kenji; Yato, Tadao; Kaneko, Hiromitsu; Honda, Yutaka

1980-01-01

Cesium seems to be one of the most important fission products in the fuel-cladding chemical interaction of fuel pins for LMFBRs. However the FCCI under irradiation cannot always be explained by considering only cesium-oxygen system as the corrosive, since attack does not occur in the cesium-oxygen system unless oxygen potential is sufficiently high. Cesium-tellurium-oxygen system has been proposed to account for heavy cladding attack which was sometimes found in hypostoichiometric mixed oxide fuel pins. In this paper, the experiment on the reaction of liquid tellurium with stainless steel is reported. The type 316 stainless steel claddings for Monju type fuel pins were used as the test specimens. Tellurium was contained into the cladding tubes with end plugs. The temperature dependence of the attack by tellurium was examined in the range from 450 to 900 deg C for 30 min, and the heating time dependence was examined from 5 min to 200 hr at 725 deg C. An infrared lamp furnace was used for the experiment within 7 hr, and a resistance furnace for longer experiment. The character of corrosion was matrix attack, and the reaction products on the stainless steel surfaces consisted of chrome rich inner phase and iron and nickel rich outer phase. The results are reported. (Kako, I.)

Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E

Energy Technology Data Exchange (ETDEWEB)

Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1968-12-15

The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.

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.

Transitioning aluminum clad spent fuels from wet to interim dry storage

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Iyer, N.C.; Sindelar, R.L.; Peacock, H.B. Jr.

1994-01-01

The United States Department of Energy (DOE) currently owns several hundred metric tons of aluminum clad, spent nuclear fuel and target assemblies. The vast majority of these irradiated assemblies are currently stored in water basins that were designed and operated for short term fuel cooling prior to fuel reprocessing. Recent DOE decisions to severely limit the reprocessing option have significantly lengthened the time of storage, thus increasing the tendency for corrosion induced degradation of the fuel cladding and the underlying core material. The portent of continued corrosion, coupled with the age of existing wet storage facilities and the cost of continuing basin operations, including necessary upgrades to meet current facility standards, may force the DOE to transition these wet stored, aluminum clad spent fuels to interim dry storage. The facilities for interim dry storage have not been developed, partially because fuel storage requirements and specifications for acceptable fuel forms are lacking. In spite of the lack of both facilities and specifications, current plans are to dry store fuels for approximately 40 to 60 years or until firm decisions are developed for final fuel disposition. The transition of the aluminum clad fuels from wet to interim dry storage will require a sequence of drying and canning operations which will include selected fuel preparations such as vacuum drying and conditioning of the storage atmosphere. Laboratory experiments and review of the available literature have demonstrated that successful interim dry storage may also require the use of fuel and canister cleaning or rinsing techniques that preclude, or at least minimize, the potential for the accumulation of chloride and other potentially deleterious ions in the dry storage environment. This paper summarizes an evaluation of the impact of fuel transitioning techniques on the potential for corrosion induced degradation of fuel forms during interim dry storage

Demonstration of fuel resistant to pellet-cladding interaction. Phase 2. First semiannual report, January-June 1979

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1979-08-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. This is the first semiannual progress report for Phase 2 of this program (January-June 1979). Progress in the irradiation testing of barrier fuel and of unfueled barrier cladding specimens is reported

On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

International Nuclear Information System (INIS)

Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

1976-05-01

The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

Fuel-pin cladding transient failure strain criterion

International Nuclear Information System (INIS)

Bard, F.E.; Duncan, D.R.; Hunter, C.W.

1983-01-01

A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

Fuel cladding tube leak detection device

International Nuclear Information System (INIS)

Naito, Makoto.

1992-01-01

The device of the present invention can detect even a minute leakage or a continuous leakage during reactor operation. That is, the device of the present invention comprises a detector for analyzing nuclides of gases incorporated in a gas waste processing system, and a calculation device connected to the detector and detecting leakage from a fuel cladding tube by calculation for variation coefficient of long-life nuclides. By using theses devices, radioactivity contained in gases incorporated in the gas waste processing system is analyzed for the nuclides. Among the analized nuclides, if the amount of the long-life nuclides exceeds a predetermined value, it is judged as leakage of the fuel cladding tube. For example, the long-life nuclides include Xe-133. The device of the present invention can certainly detect occurrence of leakage even when it is minute or continues leakage. Accordingly, countermeasures can be taken in an early stage, thereby enabling to contribute improvement for the safety of a nuclear power plant. (I.S.)

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Mariani, R.D.

1999-01-01

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

International Nuclear Information System (INIS)

Scheglov, A.

1994-01-01

In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs

Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

Energy Technology Data Exchange (ETDEWEB)

Scheglov, A [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

1994-12-31

In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs.

Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2015-10-15

Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

Pellet-clad interaction observations in boiling water reactor fuel elements

International Nuclear Information System (INIS)

Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

1981-01-01

Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

Investigation on fuel-cladding chemical interaction in metal fuel for FBR

International Nuclear Information System (INIS)

Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki

2013-01-01

During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel (PNC-FMS) and oxide-dispersion-strengthened (ODS) steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the (Fe, Cr) 17 RE 2 matrix phase with the precipitation of the (Fe, RE, Cr) phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. (author)

Compatibility study between U-UO{sub 2} cermet fuel and T91 cladding

Energy Technology Data Exchange (ETDEWEB)

Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kaity, Santu; Khan, K.B. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sengupta, Pranesh; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2016-12-01

Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO{sub 2} cermet fuel and T91 cladding material. These diffusion couples were annealed at 923–1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope (SEM), X-ray energy dispersive spectroscope (EDS) and electron probe microanalyser (EPMA). It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U{sub 6}(Fe,Cr) and U(Fe,Cr){sub 2} intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.

Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.

1978-06-01

Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

Protection of spent aluminum-clad research reactor fuels during extended wet storage

International Nuclear Information System (INIS)

Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

2013-01-01

Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

Corrosion issues in the long term storage of aluminum-clad spent nuclear fuels

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Peacock, H.B. Jr.; Sindelar, R.L.; Iyer, N.C.

1996-01-01

Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices

Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

International Nuclear Information System (INIS)

Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

1983-01-01

To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

Interdiffusion between U-Pu-Zr fuel and HT9 cladding

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Petri, M.C.

1994-01-01

As part of systematic interdiffusion studies of fuel-cladding compatibility in the integral Fast Reactor, a solid-solid diffusion couple was assembled with U-22Pu-23 1 Zr fuel and HT9 2 cladding and annealed at 650 degrees C for 100 hours. The couple was examined for diffusion structure development using a scanning electron microscope equipped with an energy dispersive x-ray analyzer (SEM-EDX). Point-by-point and linescan analysis was used to generate composition profiles and diffusion paths. From the composition profiles, average effective interdiffusion coefficients were calculated for individual components on both sides of the Matano plane. Results from this investigation indicate that the same types of phases as would be expected from binary U-Fe, Pu-Fe, and Zr-Fe phase diagrams develop in this couple; and U and Pu are the fastest diffusing fuel components and Fe is the fastest diffusing cladding component. Compared with diffusion couples with binary (U-Zr) fuel, the addition of Pu greatly enhanced the extent of diffusion and affected the types of phases observed

An internal conical mandrel technique for fracture toughness measurements on nuclear fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Sainte Catherine, C.; Le Boulch, D.; Carassou, S. [CEA Saclay, DEN/DMN, Bldg 625 P, Gif-Sur-Yvette, F-91191 (France); Lemaignan, C. [CEA Grenoble, 17 rue des Martyrs, Grenoble, F-38054 (France); Ramasubramanian, N. [ECCATEC Inc., 92 Deburn Drive, Toronto, Ontario (Canada)

2006-07-01

An understanding of the limiting stress level for crack initiation and propagation in a fuel cladding material is a fundamental requirement for the development of water reactor clad materials. Conventional tests, in use to evaluate fracture properties, are of limited help, because they are adapted from ASTM standards designed for thick materials, which differ significantly from fuel cladding geometry (small diameter thin-walled tubing). The Internal Conical Mandrel (ICM) test described here is designed to simulate the effect of fuel pellet diametrical increase on a cladding with an existing axial through-wall crack. It consists in forcing a cone, having a tapered increase in diameter, inside the Zircaloy cladding with an initial axial crack. The aim of this work is to quantify the crack initiation and propagation criteria for fuel cladding material. The crack propagation is monitored by a video system for obtaining crack extension {delta}a. A finite-element (FE) simulation of the ICM test is performed in order to derive J integrals. A node release technique is applied during the FE simulation for crack propagation and the J-resistance curves (J-{delta}a) are generated. This paper presents the test methodology, the J computation validation, and results for cold-worked stress relieved Zircaloy-4 cladding at 20 deg. and 300 deg. C and also for Al 7050-T7651 aluminum alloy tubing at 20 deg. C. (authors)

Application of Coating Technology for Accident Tolerant Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

Nuclear fuel rod with burnable plate and pellet-clad interaction fix

International Nuclear Information System (INIS)

Boyle, R.F.

1987-01-01

This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

International Nuclear Information System (INIS)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Historical fuel reprocessing and HLW management in Idaho

International Nuclear Information System (INIS)

Knecht, D.A.; Staiger, M.D.; Christian, J.D.

1997-01-01

This article review some of the key decision points in the historical development of spent fuel reprocessing and waste management practices at the Idaho Chemical Processing Plant that have helped ICPP to successfully accomplish its mission safely and with minimal impact on the environment. Topics include ICPP reprocessing development; batch aluminum-uranium dissolution; continuous aluminum uranium dissolution; batch zirconium dissolution; batch stainless steel dissolution; semicontinuous zirconium dissolution with soluble poison; electrolytic dissolution of stainless steel-clad fuel; graphite-based rover fuel processing; fluorinel fuel processing; ICPP waste management consideration and design decisions; calcination technology development; ICPP calcination demonstration and hot operations; NWCF design, construction, and operation; HLW immobilization technology development. 80 refs., 4 figs

The Effect of Peak Temperatures and Hoop Stresses on Hydride Reorientations of Zirconium Alloy Cladding Tubes under Interim Dry Storage Condition

International Nuclear Information System (INIS)

Cha, Hyun Jin; Jang, Ki Nam; Kim, Kyu Tae

2016-01-01

In this study, the effect of peak temperatures and hoop tensile stresses on hydride reorientation in cladding was investigated. It was shown that the 250ppm-H specimens generated larger radial hydride fractions and longer radial hydrides than the 500ppm-H ones. The precipitated hydride in radial direction severely degrades mechanical properties of spent fuel rod. Hydride reorientation is related to cladding material, cladding temperature, hydrogen contents, thermal cycling, hoop stress and cooling rate. US NRC established the regulation on cladding temperature during the dry storage, which is the maximum fuel cladding temperature should not exceed 400 .deg. C for all fuel burnups under normal conditions of storage. However, if it is proved that the best estimate cladding hoop stress is equal to or less than 90MPa for the temperature limit proposed, a higher short-term temperature limit is allowed for low burnup fuel. In this study, 250ppm and 500ppm hydrogen-charged Zr-Nb alloy cladding tubes were selected to evaluate the effect of peak temperatures and hoop tensile stresses on the hydride reorientation during the dry storage. In order to evaluate threshold stresses in relation to various peak temperatures, four peak temperatures of 250, 300, 350, and 400 .deg. C and three tensile hoop stresses of 80, 100, 120MPa were selected.

FUMAC-a new model for light water reactor fuel relocation and pellet-cladding interaction

International Nuclear Information System (INIS)

Walton, L.A.; Matheson, J.E.

1984-01-01

An improved approach to the mechanical modeling of fuel rod performance is presented. Previous computer modeling has centered around a unified finite element approach with both fuel pellets and cladding being represented by ring elements. The fuel mechanical analysis code (FUMAC) departs from these approaches in two areas. The pellet model is an empirically based deterministic algorithm, while the cladding model uses both plane stress and plane strain finite elements. The work describes a semiempirical fuel cracking and fragment relocation model, which is burnup and power-level dependent. The interaction of the pellet with the cladding is treated classically. The resulting thick cylinder stresses are used in conjunction with an orthotropic creep model to predict cladding ridging. The resulting ridging compares well with experimental data for both steady-state and transient operating conditions. Future work planned includes the integration of the finite element cladding model with the pellet model and refinement of the pellet relocation and thermal models. Transient performance predictions will be emphasized

The steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Bera, P.C.; Clendening, W.R.

1978-11-01

The steady-state creep rates of as-received Zircaloy-4 fuel cladding have been determined in the α-Zr phase (940 -6 and 10 -3 s -1 were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law - Arrhenius equation, the creep rate for α-phase Zircaloy-4 is given by: epsilon sub(ss) = 2000σ sup(5.32) exp (-284 600/kT) s -1 and for the β-phase Zircaloy-4 is given by: epsilon sub(ss) = 8.1σ sup(3.79) exp (-142 300/kT) s -1 . For both the α-Zr and β-Zr phases, the activation energies for creep are in agreement with those for self-diffusion of zirconium and the rate-controlling mechanism is attributed to dislocation climb. Because of the scarcity of data, it is not possible to determine the rate equation unambiguously, nor to identify the mechanism for creep in the mixed α + β phase region. (author)

Method for the protection of the cladding tubes of fuel rods

International Nuclear Information System (INIS)

Steinberg, E.

1978-01-01

To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

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.

Future possibilities of SUSEN technologies for R&D of nuclear fuel cladding

International Nuclear Information System (INIS)

Mikloš, M.

2015-01-01

R&D possibilities with nuclear fuel cladding were discussed in this paper. The availability of 10 MWT reactor with BWR and PWR loops having chemistry control was described. Activity transport and fuel cladding corrosion can be investigated in this facility including PIE. The facility has hot cells and the laboratory is expected to start in 2017

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)

Nuclear fuel element

International Nuclear Information System (INIS)

Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

1980-01-01

Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

Initial results for electrochemical dissolution of spent EBR-II fuel

International Nuclear Information System (INIS)

Li, S. X.

1998-01-01

Initial results are reported for the anode behavior of spent metallic nuclear fuel in an electrorefining process. The anode behavior has been characterized in terms of the initial spent fuel composition and the final composition of the residual cladding hulls. A variety of results have been obtained depending on the experimental conditions. Some of the process variables considered are average and maximum cell voltage, average and maximum anode voltage, amount of electrical charge passed (coulombs or amp-hours) during the experiment, and cell resistance. The main goal of the experiments has been the nearly complete dissolution of uranium with the retention of zirconium and noble metal fission products in the cladding hulls. Analysis has shown that the most indicative parameters for determining an endpoint to the process, recognizing the stated goal, are the maximum anode voltage and the amount of electrical charge passed. For the initial experiments reported here, the best result obtained is greater than 98% uranium dissolution with approximately 50% zirconium retention. Noble metal fission product retention appears to be correlated with zirconium retention

Progress in Understanding of Fuel-Cladding Chemical interaction in Metal Fuel

International Nuclear Information System (INIS)

Inagaki, Okenta; Nakamura, Kinya; Ogata, Takanari

2013-01-01

Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding (Fe). Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels

Interim report on the creepdown of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Hobson, D.O.; Dodd, C.V.

1977-01-01

This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

International Nuclear Information System (INIS)

Smith, E.; Ranjan, G.V.; Cipolla, R.C.

1976-11-01

Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

Crack resistance curves determination of tube cladding material

Energy Technology Data Exchange (ETDEWEB)

Bertsch, J. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)]. E-mail: johannes.bertsch@psi.ch; Hoffelner, W. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

2006-06-30

Zirconium based alloys have been in use as fuel cladding material in light water reactors since many years. As claddings change their mechanical properties during service, it is essential for the assessment of mechanical integrity to provide parameters for potential rupture behaviour. Usually, fracture mechanics parameters like the fracture toughness K {sub IC} or, for high plastic strains, the J-integral based elastic-plastic fracture toughness J {sub IC} are employed. In claddings with a very small wall thickness the determination of toughness needs the extension of the J-concept beyond limits of standards. In the paper a new method based on the traditional J approach is presented. Crack resistance curves (J-R curves) were created for unirradiated thin walled Zircaloy-4 and aluminium cladding tube pieces at room temperature using the single sample method. The procedure of creating sharp fatigue starter cracks with respect to optical recording was optimized. It is shown that the chosen test method is appropriate for the determination of complete J-R curves including the values J {sub 0.2} (J at 0.2 mm crack length), J {sub m} (J corresponding to the maximum load) and the slope of the curve.

Standard recommended practice for examination of fuel element cladding including the determination of the mechanical properties