Recent progress in the development of metallic fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

1990-01-01

Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

Pyroprocessing of IFR Metal Fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle features the use of an innovative reprocessing method, known as open-quotes pyroprocessingclose quotes featuring fused-salt electrofining of the spent fuel. Electrofining of IFR spent fuel involves uranium recovery by electro-transport to a solid steel cathode. The thermodynamics of the system preclude plutonium recovery in the same way, so a liquid cadmium cathode located in the electrolyte salt phase is utilized. The deposition of Pu, Am, Np, and Cm takes place at the liquid cadmium cathode in the form of cadmium intermetallic compounds (e.g, PuCd 6 ), and uranium deposits as the pure metal when cadmium saturation is reached. A small amount of rare earth fission products deposit together with the heavy metals at both the solid and liquid cadmium cathodes, providing a significant degree of self-protection. A full scope demonstration of the IFR fuel cycle will begin in 1993, using fuel irradiated in EBR-II

Advanced breeder cycle uses metallic fuel

International Nuclear Information System (INIS)

Chang, Y.I.

1991-01-01

Scientists from Argonne National Laboratory have been developing a concept called the Integral fast Reactor (IFR). This fast breeder reactor could effectively increase Uranium resources a hundred fold making nuclear power essentially an inexhaustible energy source. The IFR is outlined. In the IFR, the inherent properties of liquid metal cooling are combined with a new metallic fuel which is allowed to swell and gives an improved burnup level and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics and waste management. (author)

A state of the art on metallic fuel technology development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

Fuel motion in overpower tests of metallic integral fast reactor fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

1992-01-01

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

A state of the art on metallic fuel technology development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

Casting of metallic fuel containing minor actinide additions

International Nuclear Information System (INIS)

Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

1992-01-01

A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

Energy Technology Data Exchange (ETDEWEB)

Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

Metallic fuel design development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

International Nuclear Information System (INIS)

Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

1991-07-01

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

Development of Metallic Fuels for Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

2015-09-01

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

Non-noble metal fuel cell catalysts

CERN Document Server

Chen, Zhongwei; Zhang, Jiujun

2014-01-01

Written and edited by a group of top scientists and engineers in the field of fuel cell catalysts from both industry and academia, this book provides a complete overview of this hot topic. It covers the synthesis, characterization, activity validation and modeling of different non-noble metal and metalfree electrocatalysts for the reduction of oxygen, as well as their integration into acid or alkaline polymer exchange membrane (PEM) fuel cells and their performance validation, while also discussing those factors that will drive fuel cell commercialization. With its well-structured app

Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

International Nuclear Information System (INIS)

Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

Design study on metal fuel FBR cores

International Nuclear Information System (INIS)

Yokoo, T.; Tanaka, Y.; Ogata, T.

1991-01-01

A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

Benchmark physics tests in the metallic-fueled assembly ZPPR-15

International Nuclear Information System (INIS)

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1989-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size liquid-metal reactor (LMR) are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics, and spectrum. Analysis was done with three-dimensional nodal diffusion calculations and ENDF/B-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

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

Safeguards operations in the integral fast reactor fuel cycle

International Nuclear Information System (INIS)

Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

1994-01-01

Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

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

Fuel damage during off-normal transients in metal-fueled fast reactors

International Nuclear Information System (INIS)

Kramer, J.M.; Bauer, T.H.

1990-01-01

Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

The characterization and monitoring of metallic fuel breaches in EBR-2

International Nuclear Information System (INIS)

Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

1991-01-01

This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

A status report on the integral fast reactor fuels and safety program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

Metal fuel safety performance

International Nuclear Information System (INIS)

Miles, K.J. Jr.; Tentner, A.M.

1988-01-01

The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.

1993-01-01

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations

Simulation of the injection casting of metallic fuels

International Nuclear Information System (INIS)

Nakagawa, Tomokazu; Ogata, Takanari; Tokiwai, Moriyasu.

1989-01-01

For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

Metallic Reactor Fuel Fabrication for SFR

Energy Technology Data Exchange (ETDEWEB)

Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

1994-01-01

Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.; Kalimullah; Miles, K.J.

1990-01-01

The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities

Development of 4S and related technologies (2). Long life metallic fuel

International Nuclear Information System (INIS)

Yacout, A.M.; Tsuboi, Y.; Ueda, N.

2009-01-01

This paper provides an overview of the long life metallic fuel to be used in the 4S reactor. The 4S fuel design is presented and implications of its characteristics on fuel performance are discussed. Main design characteristics include the long fuel life time of 30 years and the wider and longer fuel pins compared to EBR-II and FFTF fuel pins. The LIFE-METAL fuel performance code was used to evaluate the performance of the 4S fuel design. The code has been validated using post irradiation examination data of metallic fuel irradiated in EBR-II. The performance evaluation shows the benign nature of the design. The design enables the fuel to perform adequately during reactor operations without violating any of a conservative set of steady state design criteria. A survey evaluation of the fuel performance is also presented. This performance bounding evaluation took into account possible fuel swelling behavior and cladding temperature range that represents worst case scenarios. The evaluation showed that the fuel maintains its integrity even under those worst case conditions. (author)

An approach for evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems

International Nuclear Information System (INIS)

Nakae, Nobuo; Ozawa, Takayuki; Ohta, Hirokazu; Ogata, Takanari; Sekimoto, Hiroshi

2014-01-01

One of the important issues in the study of Innovative Nuclear Energy Systems is evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems. An approach for evaluating the integrity of the fuel is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes determining fuel life time were reviewed and fuel integrity was analyzed and compared with the failure criteria. Metal and nitride fuels with austenitic and ferritic stainless steel (SS) cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for Innovative Nuclear Energy Systems, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic–martensitic steel (PNC–FMS) and oxide dispersion strengthened steel (PNC–ODS). The analysis showed that the fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In case of ferritic steel, on the other hand, the fuel lifetime is controlled by cladding creep rupture. The lifetime evaluated here is limited to 200 GW d/t, which is lower than the target burnup value of 500 GW d/t. One of the possible measures to extend the lifetime may be reducing the fuel smeared density and ventilating fission gas in the plenum for metal fuel and by reducing the maximum cladding temperature from 650 to 600 °C for both metal and nitride fuel

An approach for evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems

Energy Technology Data Exchange (ETDEWEB)

Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Center for Research into Innovative Nuclear Energy System, Tokyo Institute of Technology, 2-12-1-N1-19, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Ozawa, Takayuki [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4-33, Muramatsu, Tokai-mura, Ibaraki-ken 319-1194 (Japan); Ohta, Hirokazu; Ogata, Takanari [Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, 2-11-1, Iwado Kita, Komae-shi, Tokyo 201-8511 (Japan); Sekimoto, Hiroshi [Center for Research into Innovative Nuclear Energy System, Tokyo Institute of Technology, 2-12-1-N1-19, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2014-03-15

One of the important issues in the study of Innovative Nuclear Energy Systems is evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems. An approach for evaluating the integrity of the fuel is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes determining fuel life time were reviewed and fuel integrity was analyzed and compared with the failure criteria. Metal and nitride fuels with austenitic and ferritic stainless steel (SS) cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for Innovative Nuclear Energy Systems, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic–martensitic steel (PNC–FMS) and oxide dispersion strengthened steel (PNC–ODS). The analysis showed that the fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In case of ferritic steel, on the other hand, the fuel lifetime is controlled by cladding creep rupture. The lifetime evaluated here is limited to 200 GW d/t, which is lower than the target burnup value of 500 GW d/t. One of the possible measures to extend the lifetime may be reducing the fuel smeared density and ventilating fission gas in the plenum for metal fuel and by reducing the maximum cladding temperature from 650 to 600 °C for both metal and nitride fuel.

In-pile measurement of the thermal conductivity of irradiated metallic fuel

International Nuclear Information System (INIS)

Bauer, T.H.; Holland, J.W.

1995-01-01

Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

Characterization of IFR metal fuel fragmentation

International Nuclear Information System (INIS)

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

1987-01-01

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

Studies of Lanthanide Transport in Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jinsuo; Taylor, Christopher

2018-04-02

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity. The lanthanide interaction with clad in metallic fuels is recognized as a long-term, high-burnup cause of the clad failures. Therefore, one of the key concerns of using metallic fuels is the redistribution of lanthanide fission products and migration to the fuel surface. It is believed that lanthanide migration is in part due to the thermal gradient between the center and the fuel-cladding interface, but also largely in part due to the low solubility of lanthanides within the uranium-based metal fuel. PIE of EBR-II fuels shows that lanthanides precipitate directly and do not dissolve to an appreciable extent in the fuel matrix. Based on the PIE data from EBR-II, a recent study recommended a so-called “liquid-like” transport mechanism for lanthanides and certain other species. The liquid-like transport model readily accounts for redistribution of Ln, noble metal fission products, and cladding components in the fuel matrix. According to the novel mechanism, fission products can transport as solutes in liquid metals, such as liquid cesium or liquid cesium–sodium, and on pore surfaces and fracture surfaces for metals near their melting temperatures. Transport in such solutions is expected to be much more rapid than solid-state diffusion. The mechanism could explain the Ln migration to the fuel slug peripheral surface and their deposition with a sludge-like form. Lanthanides have high solubility in liquid cesium but have low solubility in liquid sodium. As a

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

International Nuclear Information System (INIS)

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

1998-10-01

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

Development of integrated metallic container

International Nuclear Information System (INIS)

Teper, B.

1985-01-01

Several alternative designs of the integrated metallic container are being investigated in an effort to obtain an optimal integrated system for transportation, storage and disposal of used fuel from CANDU Nuclear Generating Stations. Three specific designs having thin-wall, medium and thick-walls are being proposed in more detail. A novel, simplified closure system is introduced. Some economic consideration is also presented. Many ideas presented in the report deviate from the currently pursued methodology. They are presented here as an alternative for future considerations

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

Energy Technology Data Exchange (ETDEWEB)

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

2007-02-07

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

33 CFR 183.562 - Metallic fuel lines.

Science.gov (United States)

2010-07-01

...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.562 Metallic fuel lines. (a) Each metallic fuel line that is mounted to the boat structure must be connected to the engine by a flexible fuel line. (b) Each metallic fuel line must be attached to the boat's structure...

Hydrogen and fuel cell research: Institute for Integrated Energy Systems (IESVic)

International Nuclear Information System (INIS)

Pitt, L.

2006-01-01

Vision: IESVic's mission is to chart feasible paths to sustainable energy. Current research areas of investigation: 1. Energy system analysis 2. Computational fuel cell engineering; Fuel cell parameter measurement; Microscale fuel cells 3. Hydrogen dispersion studies for safety codes 4. Active magnetic refrigeration for hydrogen liquifaction and heat transfer in metal hydrides 5. Hydrogen and fuel cell system integration (author)

Apparatus for injection casting metallic nuclear energy fuel rods

Science.gov (United States)

Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

1991-01-01

Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

The analysis of fuel constituent redistribution for ternary metallic fuel slug

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Lee, Dong Uk; Kim, Young Kyun; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2004-02-01

U-TRU-Zr metallic alloy is being considered as the fuel slug for the proliferation resistance core of KALIMER. The radial fuel constituent migration is a general phenomenon in the metallic alloys. This phenomenon may affect the in-reactor performance of metallic fuel rods, influencing such factors as melting temperature, thermal conductivity, power generation rate, phase boundaries and eutectic melting of the fuel slug. Thus, constituent redistribution modeling is essential when developing a metallic fuel performance code. The constituent migration model adopted in this report was based on the Ishida's model and Hofman's theory. A subroutine program has been made and installed into the MACSIS code to simulate constituent redistribution. The radial profile of Zr redistribution was calculated for the ternary metallic fuel, and compared with the measured data.

Status of SFR Metal Fuel Development

International Nuclear Information System (INIS)

Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

2013-01-01

Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

1993-01-01

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

Integral Fast Reactor fuel pin processor

International Nuclear Information System (INIS)

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

Actinide recycle potential in the Integral Fast Reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

Chang, Y.I.; Till, C.E.

1990-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. In the electrorefining operation, uranium and plutonium are selectively transported from an anode to a cathode, leaving impurity elements, mainly fission products, either in the anode compartment or in a molten salt electrolyte. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management, because these actinides are automatically recycled back into the reactor for in-situ burning. Based on the recent IFR process development, a preliminary assessment has also been made to investigate the feasibility of further adapting the pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 5 refs., 4 figs., 4 tabs

Metal membrane-type 25-kW methanol fuel processor for fuel-cell hybrid vehicle

Science.gov (United States)

Han, Jaesung; Lee, Seok-Min; Chang, Hyuksang

A 25-kW on-board methanol fuel processor has been developed. It consists of a methanol steam reformer, which converts methanol to hydrogen-rich gas mixture, and two metal membrane modules, which clean-up the gas mixture to high-purity hydrogen. It produces hydrogen at rates up to 25 N m 3/h and the purity of the product hydrogen is over 99.9995% with a CO content of less than 1 ppm. In this fuel processor, the operating condition of the reformer and the metal membrane modules is nearly the same, so that operation is simple and the overall system construction is compact by eliminating the extensive temperature control of the intermediate gas streams. The recovery of hydrogen in the metal membrane units is maintained at 70-75% by the control of the pressure in the system, and the remaining 25-30% hydrogen is recycled to a catalytic combustion zone to supply heat for the methanol steam-reforming reaction. The thermal efficiency of the fuel processor is about 75% and the inlet air pressure is as low as 4 psi. The fuel processor is currently being integrated with 25-kW polymer electrolyte membrane fuel-cell (PEMFC) stack developed by the Hyundai Motor Company. The stack exhibits the same performance as those with pure hydrogen, which proves that the maximum power output as well as the minimum stack degradation is possible with this fuel processor. This fuel-cell 'engine' is to be installed in a hybrid passenger vehicle for road testing.

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

Waste management in IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.; Battles, J.E.

1991-01-01

The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

International Nuclear Information System (INIS)

Tanaka, Nobuo

2017-01-01

Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

Influence of the resonance integral value on the fuel cycle characteristics

Energy Technology Data Exchange (ETDEWEB)

Graziani, G; Trauwaert, E

1972-04-24

The problem that is considered here is to determine what can be done about a variation in resonance integral when the complete geometry of the reactor and of the fuel elements are fixed, leaving as only free parameters the amount of heavy metal and the enrichment to put in the fuel pins.

Development of metal cask for nuclear spent fuel

International Nuclear Information System (INIS)

Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

2001-01-01

It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-01-01

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

International Nuclear Information System (INIS)

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

Evaluation report(1): on design criteria for KALIMER metal fuel

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung Oon; Kim, Young Il

2001-04-01

Fuel rods, assembly ducts and their components in KALIMER should be designed to maintain the integrities and to assure their reliable in-reactor performances under the steady state and operational transient conditions which are included in design basis category. And the fuel system must be designed with enough engineering margin to minimize and prevent the failures under ab-normal operational condition, like an accident.In this report, some design limits and the criteria for the fuel assembly ducts for KALIMER are driven by evaluating the irradiation data of metallic fuel based on experimental data from ANL in USA, CRIEPI in Japan and RIAR in Russia

Evaluation report(1): on design criteria for KALIMER metal fuel

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Lee, Byoung Oon; Kim, Young Il

2001-04-01

Fuel rods, assembly ducts and their components in KALIMER should be designed to maintain the integrities and to assure their reliable in-reactor performances under the steady state and operational transient conditions which are included in design basis category. And the fuel system must be designed with enough engineering margin to minimize and prevent the failures under ab-normal operational condition, like an accident.In this report, some design limits and the criteria for the fuel assembly ducts for KALIMER are driven by evaluating the irradiation data of metallic fuel based on experimental data from ANL in USA, CRIEPI in Japan and RIAR in Russia.

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

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

Development of metallic fuel fabrication

Energy Technology Data Exchange (ETDEWEB)

Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

1999-03-01

With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

Proposed fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Walters, L.C.

1985-01-01

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

Alternative Fabrication of Recycling Fast Reactor Metal Fuel

International Nuclear Information System (INIS)

Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

2015-01-01

Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

1986-01-01

In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

Actinide recycle potential in the integral fast reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

Chang, Y.I.; Till, C.E.

1991-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Based on the recent IFR process development, a preliminary assessment has been made to investigate the feasibility of further adapting pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs

Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor

International Nuclear Information System (INIS)

Burris, L.; Steindler, M.; Miller, W.

1984-01-01

The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

Validation of the metal fuel version of the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.

1991-01-01

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

Integral nuclear fuel element assembly

International Nuclear Information System (INIS)

Schluderberg, D. C.

1985-01-01

An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Thermal Expansion Property of U-Zr Alloys and U-Zr-Ce Alloys as a Surrogate Metallic Fuel for SFR

International Nuclear Information System (INIS)

Kim, Sun Ki; Lee, Jong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Youn Myung; Lee, Chan Bock

2010-01-01

Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. An extensive database on the performance of advanced metal fuels was generated as a result of the operation of these reactors and the IFR program. In this study, the U-Zr binary alloys and U-Zr-Ce ternary alloys as surrogate metallic fuel were fabricated in lower pressure Ar environment by gravity casting. The melt temperature was approximately 1,500 .deg. C. Thermal expansion of the fuel during normal operation is related with fuel performance in a reactor. Therefore, it is necessary to investigate the thermal expansion of the fuel in order to warrant a good prediction the fuel performance

Preparations for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.

1989-01-01

Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

Proliferation resistance of the fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.

1993-01-01

Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection

Evaluation of Metal-Fueled Surface Reactor Concepts

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

2007-01-01

Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

Optimization of Additive-Powder Characteristics for Metallic Micro-Cell UO{sub 2} Fuel Pellet Fabrication

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various aspects. The mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capability of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. The nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed. Typically, an enhancement of the thermal conductivity of the UO{sub 2} fuel pellet can be obtained by the addition of a higher thermal conductive material in the fuel pellet. To maximize the effect of the thermal conductivity enhancement, a continuous and uniform channel of the thermal conductive material in the UO{sub 2} matrix must be formed. To enhance the thermal conductivity of a UO{sub 2} fuel pellet, the development of fabrication process of a Cr metallic micro-cell UO{sub 2} pellet with a continuous and uniform channel of the Cr metallic phase was carried out. The formation of the Cr-oxide phases was prevented and the uniformity of the Cr-metal phase distribution was enhanced simultaneously, through the optimization of the additive-powder characteristics. In the results, the Cr metallic micro-cell pellet with continuous and uniform Cr metallic channel could be obtained.

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

1993-01-01

Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Steindler, M.; Miller, W.

1984-01-01

The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

Performance tests for integral reactor nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

2006-02-15

An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

International Nuclear Information System (INIS)

Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

1989-01-01

The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

Weld Joint Design for SFR Metallic Fuel Element Closures

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

International Nuclear Information System (INIS)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

2013-01-01

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

46 CFR 119.435 - Integral fuel tanks.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Integral fuel tanks. 119.435 Section 119.435 Shipping... Machinery Requirements § 119.435 Integral fuel tanks. (a) Diesel fuel tanks may not be built integral with... for certification of a vessel, integral fuel tanks must withstand a hydrostatic pressure test of 35 k...

Progress and status of the integral fast reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

Metal-deactivating additives for liquid fuels

Energy Technology Data Exchange (ETDEWEB)

Boneva, M.I. [Institute of Organic Chemistry, Sofia (Bulgaria); Ivanov, S.K.; Kalitchin, Z.D. [SciBulCom, Ltd., Sofia (Bulgaria); Tanielyan, S.K. [Seton Hall Univ., South Orange, NJ (United States); Terebenina, A.; Todorova, O.I. [Institute of Inorganic Chemistry, Sofia (Bulgaria)

1995-05-01

The metal-deactivating and the antioxidant properties of 1-phenyl-3-methylpyrazolone-5 derivatives have been investigated both in the model reaction of low temperature oxidation of ethylbenzene and in gasoline oxidation. The study of the ability of these derivatives to reduce the catalytic effect of copper naphthenate demonstrates that they are promising as metal deactivating additives for light fuels. Some of the pyrazolone compounds appear to be of special interest for the long-term storage of liquid fuels due to their action as multifunctional inhibitors.

Fabrication of metallic fuel for fast breeder reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

2012-01-01

Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

International Nuclear Information System (INIS)

Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

2010-01-01

The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

Vertical integration in the nuclear fuel cycle

International Nuclear Information System (INIS)

Mommsen, J.T.

1977-01-01

Vertical integration in the nuclear fuel cycle and its contribution to market power of integrated fuel suppliers were studied. The industry subdivision analyzed is the uranium raw materials sector. The hypotheses demonstrated are that (1) this sector of the industry is trending toward vertical integration between production of uranium raw materials and the manufacture of nuclear fuel elements, and (2) this vertical integration confers upon integrated firms a significant market advantage over non-integrated fuel manufacturers. Under microeconomic concepts the rationale for vertical integration is the pursuit of efficiency, and it is beneficial because it increases physical output and decreases price. The Market Advantage Model developed is an arithmetical statement of the relative market power (in terms of price) between non-integrated nuclear fuel manufacturers and integrated raw material/fuel suppliers, based on the concept of the ''squeeze.'' In operation, the model compares net profit and return on sales of nuclear fuel elements between the competitors, under different price and cost circumstances. The model shows that, if integrated and non-integrated competitors sell their final product at identical prices, the non-integrated manufacturer returns a net profit only 17% of the integrated firm. Also, the integrated supplier can price his product 35% below the non-integrated producer's price and still return the same net profit. Vertical integration confers a definite market advantage to the integrated supplier, and the basic source of that advantage is the cost-price differential of the raw material, uranium

Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

46 CFR 182.435 - Integral fuel tanks.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Integral fuel tanks. 182.435 Section 182.435 Shipping...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.435 Integral fuel tanks. (a) Gasoline fuel tanks must be independent of the hull. (b) Diesel fuel tanks may not be built integral with the hull of...

Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

International Nuclear Information System (INIS)

Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

1999-02-01

Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

Fabrication of particulate metal fuel for fast burner reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

2012-01-01

U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

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)

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

William J. Carmack

2012-05-01

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

Mesoporous yttria-zirconia and metal-yttria-zirconia solid solutions for fuel cells

Energy Technology Data Exchange (ETDEWEB)

Mamak, M.; Coombs, N.; Ozin, G. [Toronto Univ., ON (Canada). Dept. of Chemistry

2000-02-03

A new class of binary mesoporous yttria-zirconia (YZ) and ternary mesoporous metal-YZ materials (M = electroactive Ni/Pt) is presented here that displays the highest surface area of any known form of yttria-stabilized zirconia. These mesoporous materials form as solid solutions and retain their structural integrity to 800 C, which bodes well for their possible utilization in fuel cells. (orig.)

Long term integrity of spent fuel and construction materials for dry storage facilities

Energy Technology Data Exchange (ETDEWEB)

Saegusa, T [CRIEPI (Japan)

2012-07-01

In Japan, two dry storage facilities at reactor sites have already been operating since 1995 and 2002, respectively. Additionally, a large scale dry storage facility away from reactor sites is under safety examination for license near the coast and desired to start its operation in 2010. Its final storage capacity is 5,000tU. It is therefore necessary to obtain and evaluate the related data on integrity of spent fuels loaded into and construction materials of casks during long term dry storage. The objectives are: - Spent fuel rod: To evaluate hydrogen migration along axial fuel direction on irradiated claddings stored for twenty years in air; To evaluate pellet oxidation behaviour for high burn-up UO{sub 2} fuels; - Construction materials for dry storage facilities: To evaluate long term reliability of welded stainless steel canister under stress corrosion cracking (SCC) environment; To evaluate long term integrity of concrete cask under carbonation and salt attack environment; To evaluate integrity of sealability of metal gasket under long term storage and short term accidental impact force.

Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

1990-01-01

Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

Low Loss Advanced Metallic Fuel Casting Evaluation

International Nuclear Information System (INIS)

Kim, Kihwan; Ko, Youngmo; Kim, Jonghwan; Song, Hoon; Lee Chanbock

2014-01-01

The fabrication process for SFR fuel is composed of fuel slug casting, loading and fabrication of the fuel rods, and the fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycles streams in the fabrication process. Recycle streams include fuel slug reworks, returned scraps, and fuel casting heels, which are a special concern in the counter gravity injection casting process because of the large masses involved. Large recycle and waste streams result in lowering the productivity and the economic efficiency of fuel production. To increase efficiency the fuel losses in the furnace chamber, crucible, and the mold, after casting a considerable amount of fuel alloy in the casting furnace, will be quantitatively evaluated. After evaluation the losses will be identified and minimized. It is expected that this study will contribute to the minimization of fuel losses and the wastes streams in the fabrication process of the fuel slugs. Also through this study the technical readiness level of the metallic fuel fabrication process will be further enhanced. In this study, U-Zr alloy system fuel slugs were fabricated by a gravity casting method. Metallic fuel slugs were successfully fabricated with 19 slugs/batch with diameter of 5mm and length of 300mm. Fuel losses was quantitatively evaluated in casting process for the fuel slugs. Fuel losses of the fuel slugs were so low, 0.1∼1.0%. Injection casting experiments have been performed to reduce the fuel loss and improve the casting method. U-Zr fuel slug having φ5.4-L250mm was soundly fabricated with 0.1% in fuel loss. The fuel losses could be minimized to 0.1%, which showed that casting technology of fuel slugs can be a feasible approach to reach the goal of the fuel losses of 0.1% or less in commercial scale

Metal-fuel modeling for inherently safe reactor designs

International Nuclear Information System (INIS)

Miles, K.J. Jr.

1987-01-01

Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

Integration of the AVLIS [atomic vapor laser isotopic separation] process into the nuclear fuel cycle

International Nuclear Information System (INIS)

Hargrove, R.S.; Knighton, J.B.; Eby, R.S.; Pashley, J.H.; Norman, R.E.

1986-08-01

AVLIS RD and D efforts are currently proceeding toward full-scale integrated enrichment demonstrations in the late 1980's and potential plant deployment in the mid 1990's. Since AVLIS requires a uranium metal feed and produces an enriched uranium metal product, some change in current uranium processing practices are necessitated. AVLIS could operate with a UF 6 -in UF 6 -out interface with little effect to the remainder of the fuel cycle. This path, however, does not allow electric utility customers to realize the full potential of low cost AVLIS enrichment. Several alternative processing methods have been identified and evaluated which appear to provide opportunities to make substantial cost savings in the overall fuel cycle. These alternatives involve varying levels of RD and D resources, calendar time, and technical risk to implement and provide these cost reduction opportunities. Both feed conversion contracts and fuel fabricator contracts are long-term entities. Because of these factors, it is not too early to start planning and making decisions on the most advantageous options so that AVLIS can be integrated cost effectively into the fuel cycle. This should offer economic opportunity to all parties involved including DOE, utilities, feed converters, and fuel fabricators. 10 refs., 11 figs., 2 tabs

Integration of the AVLIS (atomic vapor laser isotopic separation) process into the nuclear fuel cycle. [Effect of AVLIS feed requirements on overall fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Hargrove, R.S.; Knighton, J.B.; Eby, R.S.; Pashley, J.H.; Norman, R.E.

1986-08-01

AVLIS RD and D efforts are currently proceeding toward full-scale integrated enrichment demonstrations in the late 1980's and potential plant deployment in the mid 1990's. Since AVLIS requires a uranium metal feed and produces an enriched uranium metal product, some change in current uranium processing practices are necessitated. AVLIS could operate with a UF/sub 6/-in UF/sub 6/-out interface with little effect to the remainder of the fuel cycle. This path, however, does not allow electric utility customers to realize the full potential of low cost AVLIS enrichment. Several alternative processing methods have been identified and evaluated which appear to provide opportunities to make substantial cost savings in the overall fuel cycle. These alternatives involve varying levels of RD and D resources, calendar time, and technical risk to implement and provide these cost reduction opportunities. Both feed conversion contracts and fuel fabricator contracts are long-term entities. Because of these factors, it is not too early to start planning and making decisions on the most advantageous options so that AVLIS can be integrated cost effectively into the fuel cycle. This should offer economic opportunity to all parties involved including DOE, utilities, feed converters, and fuel fabricators. 10 refs., 11 figs., 2 tabs.

Structure of fuel performance audit code for SFR metal fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-10-15

A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

New Nuclear Materials Including Non Metallic Fuel Elements. Vol. I. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

International Nuclear Information System (INIS)

1963-01-01

One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers

Behavior of actinides in the Integral Fast Reactor fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

1994-06-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Behavior of actinides in the Integral Fast Reactor fuel cycle

International Nuclear Information System (INIS)

Courtney, J.C.; Lineberry, M.J.

1994-01-01

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

2013-05-15

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

International Nuclear Information System (INIS)

Karahan, Aydın; Andrews, Nathan C.

2013-01-01

Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

Behavior of metallic fuel in treat transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

1988-01-01

Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

A basic research on the transient behavior for a metallic fuel FBR

International Nuclear Information System (INIS)

Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

1999-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

Design characteristics of metallic fuel rod on its in-LMR performance

International Nuclear Information System (INIS)

Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

International Nuclear Information System (INIS)

Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

2012-01-01

This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

International Nuclear Information System (INIS)

Yang, W.S.; Kim, T.K.; Grandy, C.

2007-01-01

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

The integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Yoon I.; Marchaterre, J.F.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

International Nuclear Information System (INIS)

Travelli, A.

1988-01-01

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

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

The basic research on the CDA initiation phase for a metallic fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

1998-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

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)

Preliminary study on metallic inclusion in nuclear fuel

International Nuclear Information System (INIS)

Fukuzawa, Takashi; Tanaka, Masahiro; Tanabe, Tetsuo; Imoto, Shosuke

1984-01-01

In recent postirradiation tests, metallic fission products such as Mo, Ru, Rh and Pd are known to precipitate as metallic inclusions in the fuel. These inclusions remain as insoluble residues and provide various problems in different fields of the reprocessing. In this report are presented preliminary results of the study on the ternary phase diagram of Mo-Ru-Pd system and on their properties in nitric acid or various oxidative environments. It is concluded that (1) most of metallic inclusions which are insoluble in nitric acid show epsilon phase, Ru base hcp alloy, in which a large amount of Mo and Pd are soluble, (2) Pd, however, seems to deposit separately in the fuel pin because of its high vapor pressure, (3) Mo fraction in the inclusion would be highly dependent on oxygen potential in the fuel pin. (author)

Metal waste forms from treatment of EBR-II spent fuel

International Nuclear Information System (INIS)

Abraham, D. P.

1998-01-01

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

Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

1993-01-01

The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

Electrorefining open-quotes Nclose quotes reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.

1995-01-01

Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined

Advancing liquid metal reactor technology with nitride fuels

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

1991-08-01

A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

International Nuclear Information System (INIS)

Siefken, L.J.

1999-01-01

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown

Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

International Nuclear Information System (INIS)

Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

2007-01-01

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

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

Contribution of metallic fission product inclusions to axial fuel motion potential

International Nuclear Information System (INIS)

Sasa, P.; Cronenberg, A.; Stevenson, M.

1979-01-01

In the analysis of postulated nuclear reactor accidents, axial fuel motion within the fuel pin prior to cladding failure can have an important mitigating effect. The question of primary importance is whether or not metallic inclusions have the potential to vaporize during an overheating event and thus contribute to fuel motion. To assess this potential, two limiting calculations were made: 1) The inclusion constituent assumed insoluble in one another and 2) The constituents assumed totally miscible in one another. Thermodynamic considerations indicate that the metallic fission products found within inclusions of fuel rods irradiated in a fast neutron spectrum, would form homogeneous solutions. Therefore, it is concluded that the metallic fission products would not enhance fuel swelling during an overheating event. 16 refs

A High Integrity Can Design for Degraded Nuclear Fuel

International Nuclear Information System (INIS)

Holmes, P.A.

1999-01-01

A high integrity can (HIC), designed to meet the ASME Boiler and Pressure Vessel Code (Section III, Div. 3, static conditions) is proposed for the interim storage and repository disposal of Department of Energy (DOE) spent nuclear fuel. The HIC will be approximately 5 3/8 inches (134.38mm) in outside diameter with 1/4 inch (6.35mm) thick walls, and have a removable lid with a metallic seal that is capable of being welded shut. The opening of the can is approximately 4 3/8 inches (111.13mm). The HIC is primarily designed to contain items in the DOE SNF inventory that do not meet acceptance standards for direct disposal in a geologic repository. This includes fuel in the form of particulate dusts, sectioned pieces of fuel, core rubble, melted or degraded (non-intact) fuel elements, unclad uranium alloys, metallurgical specimens, and chemically reactive fuel components. The HIC is intended to act as a substitute cladding for the spent nuclear fuel, further isolate problematic materials, provide a long-term corrosion barrier, and add an extra internal pressure barrier to the waste package. The HIC will also delay potential fission product release and maintain geometry control for extended periods of time. For the entire disposal package to be licensed by the Nuclear Regulatory Commission, a HIC must effectively eliminate the disposal problems associated with problem SNF including the release of radioactive and/or reactive material and over pressurization of the HIC due to chemical reactions within the can. Two HICs were analyzed to envelop a range of can lengths between 42 and 101 inches. Using Abacus software, the HIC's were analyzed for end, side, and corner drops. Hastelloy C-22 was chosen based upon structural integrity, corrosion resistance, and neutron adsorption properties

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

1992-01-01

Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

International Nuclear Information System (INIS)

Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

1990-01-01

In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Palomares, R.I.; Dayman, K.J.; Landsberger, S.; Biegalski, S.R.; Soderquist, C.Z.; Casella, A.J.; Brady Raap, M.C.; Schwantes, J.M.

2015-01-01

Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

A Deformation Model of TRU Metal Dispersion Fuel Rod for HYPER

International Nuclear Information System (INIS)

Lee, Byoung Oon; Hwang, Woan; Park, Won S.

2002-01-01

strain is about 3.01% at 30 at%. As the design requirement for cladding is assumed to be about 3% of cladding strain, a detailed review for cladding strain is needed. Therefore, some experimental tests including in-pile and out-pile experiments are needed for clarifying the integrity and material properties of the inter-metallic fuel core. (authors)

The evaluation of the use of metal alloy fuels in pressurized water reactors

International Nuclear Information System (INIS)

Lancaster, D.

1992-01-01

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

Integrated oil production and upgrading using molten alkali metal

Science.gov (United States)

Gordon, John Howard

2016-10-04

A method that combines the oil retorting process (or other process needed to obtain/extract heavy oil or bitumen) with the process for upgrading these materials using sodium or other alkali metals. Specifically, the shale gas or other gases that are obtained from the retorting/extraction process may be introduced into the upgrading reactor and used to upgrade the oil feedstock. Also, the solid materials obtained from the reactor may be used as a fuel source, thereby providing the heat necessary for the retorting/extraction process. Other forms of integration are also disclosed.

Design of metallic bipolar plates for PEM fuel cells.

Science.gov (United States)

2012-01-01

This project focused on the design and production of metallic bipolar plates for use in PEM fuel cells. Different metals were explored : and stainless steel was found out to be best suited to our purpose. Following the selection of metal, it was calc...

Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

International Nuclear Information System (INIS)

Orechwa, Y.; Bucher, R.G.

1994-01-01

The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software

Integrated fuel processor development

International Nuclear Information System (INIS)

Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

2001-01-01

The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

Fabrication of Metallic Fuel Slugs for Irradiation Experiments in Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arun; Prasad, G.J.

2013-01-01

Advantages of Metallic fuels for future FBR: → High heavy metal atom density; → Higher thermal conductivity at room temperature that increases with temperature; → Metal fuels can be relatively easily fabricated with close dimensional tolerances; → They have excellent compatibility with liquid metal coolants

Plans for the development of the IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.

1986-01-01

The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium

Assesment of strength and integrity of fuel channels

International Nuclear Information System (INIS)

2000-01-01

Detailed analysis to base strength and integrity of fuel channels was necessary for the licensing process. Description of tasks performed in this direction in 1999 is presented: fuel channel independent strength calculations, assessment of present fuel channels state, analysis of dynamic processes during partial group distribution header rupture, structural integrity analysis of fuel channels located next to broke channel

Dimensional, microstructural and compositional stability of metal fuels

International Nuclear Information System (INIS)

Solomon, A.A.; Dayananda, M.A.

1993-01-01

The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr

Integrated spent nuclear fuel database system

International Nuclear Information System (INIS)

Henline, S.P.; Klingler, K.G.; Schierman, B.H.

1994-01-01

The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

Cumulative damage fraction design approach for LMFBR metallic fuel elements

International Nuclear Information System (INIS)

Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

1979-01-01

The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

Energy Technology Data Exchange (ETDEWEB)

Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B

1999-03-01

The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.

Metal Matrix Microencapsulated Fuel Technology for LWR Applications

International Nuclear Information System (INIS)

Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

2012-01-01

An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

Safety characteristics of the integral fast reactor concept

International Nuclear Information System (INIS)

Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

1985-01-01

The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

Simulation of integral local tests with high-burnup fuel

International Nuclear Information System (INIS)

Gyori, G.

2011-01-01

The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

Spent nuclear fuel project integrated schedule plan

International Nuclear Information System (INIS)

Squires, K.G.

1995-01-01

The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel

Spent nuclear fuel project integrated schedule plan

Energy Technology Data Exchange (ETDEWEB)

Squires, K.G.

1995-03-06

The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

Integral fast reactor safety features

International Nuclear Information System (INIS)

Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

1988-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

Directory of Open Access Journals (Sweden)

S. K. Kim

2013-01-01

Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

International Nuclear Information System (INIS)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-01-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Development and testing of metallic fuels with high minor actinide content

International Nuclear Information System (INIS)

Meyer, M.K.; Hayes, S.L.; Kennedy, J.R.; Keiser, D.D.; Hilton, B.A.; Frank, S.M.; Kim, Y.-S.; Chang, G.; Ambrosek, R.G.

2003-01-01

Metallic alloys are promising candidates for use as fuels for transmutation and in advanced closed nuclear cycles. Metallic alloys have high heavy metal atom density, relatively high thermal conductivity, favorable gas release behavior, and lend themselves to remote recycle processes. Both non-fertile and uranium-bearing metal fuels containing minor actinide are under consideration for use as transmutation fuels by the U.S. Advanced Fuel Cycle (AFC) program, however, little irradiation performance data exists for fuel forms containing significant fractions of minor actinides. The first irradiation tests of non-fertile high-actinide-content fuels are scheduled to begin in early 2003 in the Advanced Test Reactor (ATR). The irradiation test matrix was designed to provide basic information on the irradiation behavior of binary Pu-Zr alloy fuel and the effect of the minor actinides americium and neptunium on alloy fuel behavior, together and separately. Five variants of transuranic containing zirconium-based alloy fuels are included in the AFC-1 irradiation test matrix. These are (in wt.%) Pu-40Zr, Pu-60Zr, Pu-12Am-40Zr, Pu-10Np-40Zr and Pu-10Np-10Am-40Zr. PuN-ZrN based fuels containing Am and Np are also included. All five of the fuel alloys have been fabricated in the form of cylindrical fuel slugs by arc-casting. Short melt times, on the order or 5-20 seconds, prevent the volatilization of significant quantities of americium metal, despite the high melt temperatures characteristic of the arc-melting process. Alloy microstructure have been characterized by x-ray diffraction and scanning electron microscopy. Thermal analysis has also been performed. The AFC-1 irradiation experiment configuration consists of twenty-four sodium bonded fuel specimens sealed in helium filled secondary capsules. The first capsule has a design burnup to 7 at.% 239 Pu; goal peak burnup of the second capsule is ∼18 at%. Capsule assemblies are placed within an aluminum flow-through basket

Evaluation of the use of metal alloy fuels in pressurized water reactors

International Nuclear Information System (INIS)

1990-01-01

The project concentrated on model development. Reactor physics modeling involved establishing accurate models with PC versions of COMBINE and VENTURE. Fuel performance analysis will start with METAL- LIFE. In order to justify the change of fuel to metal alloy, large benefits will have to be found; the cost benefit reported is not sufficient. The fuel pin will be annular and contact the clad; the clad thickness will force the fuel to grow toward the central hole. This report reports: design improvements, neutronic model development, COBRA modifications, reactor kinetics model development, RELAP code, and fuel performance

Response of Salix alba L. to heavy metals and diesel fuel ...

African Journals Online (AJOL)

Response of Salix alba L. to heavy metals and diesel fuel contamination. ... Ni and Pb led to reduced accumulation of Pb. Presence of 5 g/kg of diesel fuel in soil significantly increased toxic influence of applied heavy metals by further reducing ... Key words: Cd, Ni, Pb, phytoremediation, phytoextraction, willow, Salix alba.

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

Energy Technology Data Exchange (ETDEWEB)

Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2017-07-05

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

Treatment of wastes in the Integral Fast Reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.; Chow, L.S.H.; Carls, E.L.; Hannum, W.H.; Laidler, J.J.

1997-01-01

In both the reactor portion and the fuel-cycle portion of the Integral Fast Reactor (IFR), handling, treatment and disposal of wastes are simpler than in current fuel cycles. The vast majority (> 99.9%) of the very-long-lived radioactive TRU elements are not sent to the repository; rather, they are recycled. High-level waste volume from the IFR process (called ''the pyroprocess'') is lower than that from either the direct disposal of spent fuel or from conventional PUREX-type reprocessing. The quantity of low-level waste is very low. In the pyroprocess, the actinides are recovered and separated from the bulk of the fission products by an electrorefining step wherein the actinides are electrotransported from chopped fuel elements and deposited at cathodes. The volatile fission products xenon, krypton, and tritium are collected for long-term storage and decay. Zirconium and the ''noble metal'' fission products (those that are less easily oxidized than zirconium) remain in the anode compartment, to be removed with the fuel cladding fragments and made into a metal waste form. The remaining fission products collect in the salt as chlorides. A process has been developed to periodically remove the contaminated salt from the electrorefiner, separate most of the fission products, and return the purified salt in a form that is ready for continuing use. To clean up the electrorefiner salt, the fission products are removed by ion exchange onto a column of Zeolite A. After the purification step, the column material and the contained fission products are converted to a mineral waste form for disposal. The processes and equipment for waste isolation and conversion to suitable disposal forms are described in this paper. (author)

Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

Energy Technology Data Exchange (ETDEWEB)

Ziomek-Moroz, M.; Hawk, Jeffrey A.

2005-01-01

Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Integral fast reactor safety features

International Nuclear Information System (INIS)

Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

1988-01-01

The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

Integrated fuel-cycle models for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.; Maudlin, P.J.

1981-01-01

Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Electrometallurgical treatment of metallic spent nuclear fuel stored at the Hanford Site

International Nuclear Information System (INIS)

Laidler, J.J.; Gay, E.C.

1996-01-01

The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

Oxy-fuel combustion with integrated pollution control

Science.gov (United States)

Patrick, Brian R [Chicago, IL; Ochs, Thomas Lilburn [Albany, OR; Summers, Cathy Ann [Albany, OR; Oryshchyn, Danylo B [Philomath, OR; Turner, Paul Chandler [Independence, OR

2012-01-03

An oxygen fueled integrated pollutant removal and combustion system includes a combustion system and an integrated pollutant removal system. The combustion system includes a furnace having at least one burner that is configured to substantially prevent the introduction of air. An oxygen supply supplies oxygen at a predetermine purity greater than 21 percent and a carbon based fuel supply supplies a carbon based fuel. Oxygen and fuel are fed into the furnace in controlled proportion to each other and combustion is controlled to produce a flame temperature in excess of 3000 degrees F. and a flue gas stream containing CO2 and other gases. The flue gas stream is substantially void of non-fuel borne nitrogen containing combustion produced gaseous compounds. The integrated pollutant removal system includes at least one direct contact heat exchanger for bringing the flue gas into intimated contact with a cooling liquid to produce a pollutant-laden liquid stream and a stripped flue gas stream and at least one compressor for receiving and compressing the stripped flue gas stream.

The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

Energy Technology Data Exchange (ETDEWEB)

Lancaster, D.

1992-10-26

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

Metallic uranium as fuel for fast reactors

International Nuclear Information System (INIS)

Moura Neto, C. de

1988-01-01

This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

Purifier-integrated methanol reformer for fuel cell vehicles

Science.gov (United States)

Han, Jaesung; Kim, Il-soo; Choi, Keun-Sup

We developed a compact, 3-kW, purifier-integrated modular reformer which becomes the building block of full-scale 30-kW or 50-kW methanol fuel processors for fuel cell vehicles. Our proprietary technologies regarding hydrogen purification by composite metal membrane and catalytic combustion by washcoated wire-mesh catalyst were combined with the conventional methanol steam-reforming technology, resulting in higher conversion, excellent quality of product hydrogen, and better thermal efficiency than any other systems using preferential oxidation. In this system, steam reforming, hydrogen purification, and catalytic combustion all take place in a single reactor so that the whole system is compact and easy to operate. Hydrogen from the module is ultrahigh pure (99.9999% or better), hence there is no power degradation of PEMFC stack due to contamination by CO. Also, since only pure hydrogen is supplied to the anode of the PEMFC stack, 100% hydrogen utilization is possible in the stack. The module produces 2.3 Nm 3/h of hydrogen, which is equivalent to 3 kW when PEMFC has 43% efficiency. Thermal efficiency (HHV of product H 2/HHV of MeOH in) of the module is 89% and the power density of the module is 0.77 kW/l. This work was conducted in cooperation with Hyundai Motor Company in the form of a Korean national project. Currently the module is under test with an actual fuel cell stack in order to verify its performance. Sooner or later a full-scale 30-kW system will be constructed by connecting these modules in series and parallel and will serve as the fuel processor for the Korean first fuel cell hybrid vehicle.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

Energy Technology Data Exchange (ETDEWEB)

Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

International Nuclear Information System (INIS)

Stephen, Neethu Hanna; Reddy, C.P.

2013-01-01

Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

Integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

1984-01-01

Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept

Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

International Nuclear Information System (INIS)

Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

1998-01-01

The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

Fuel transfer manipulator for liquid metal nuclear reactors

International Nuclear Information System (INIS)

Sturges, R.H.

1983-01-01

A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

Production of metal waste forms from spent fuel treatment

International Nuclear Information System (INIS)

Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.

1995-01-01

Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities

Integrated planning for a fuel industry with emphasis on minimum size to fabricate own fuel

International Nuclear Information System (INIS)

Kondal Rao, N.; Katiyar, H.C.; Rajendran, R.; Sinha, K.K.; Swaminathan, N.; Subramanyam, R.B.; Pande, B.P.; Krishnan, T.S.; Agarwala, G.C.; Chandramouli, V.A.

1977-01-01

The Indian nuclear energy programme is based on the utilization of indigenous resources for the economic generation of power, developing its own know-how. In order to gain time, the first nuclear power station at Tarapur is a turn-key job based on enriched uranium fuel. Taking into consideration the established resources of uranium and thorium in the country, a strategy for nuclear power programme has been drawn up. The first phase is based on natural uranium fuel, the second phase on the recycle of plutonium and conversion of thorium and the third phase is the breeder system based on utilization of U 233 and conversion of thorium. This programme is specially significant for India in view of its vast resources of thorium. After the experience and confidence gained with the manufacture of metallic uranium fuel for the research reactors and about 40 tonnes of fuel for the initial loading of the Rajasthan Reactor, the fuel manufacturing programme within the country has been implemented to meet the entire initial and reload fuel requirements. The plant capacities are small compared to similar activities in developed countries. Further, by planning for an integrated fuel and component manufacturing complex, any draw-back in smaller scale of some of the operations is off-set. At the Nuclear Fuel Complex, set up on the above principles, production plants are in operation for the manufacture of reload fuel for the 400 MW Tarapur station, natural uranium oxide fuel, various zircaloy components such as fuel sheaths, pressure tubes, calandria tubes, channels and various other zircaloy components. Provisions have been made to expand the production facilities as the demand for reload fuel grows. With the facilities provided, the production programme can be diversified to take up the production of fast breeder reactor components of stainless steel and also the blanket thorium elements. The unitary control of all aspects of the manufacture and quality control of different types

Oxidation behaviour of noble-metal inclusions in used UO2 nuclear fuel

International Nuclear Information System (INIS)

McEachern, R.

1997-07-01

The literature on the chemistry of the noble-metal (Mo-Rh-Ru-Pd-Tc) inclusions found in used nuclear fuel has been reviewed. The Mo-Ru-Pd phase diagram is reasonably well understood, and the pseudoternary Mo-(Tc+Ru)-Rh+Pd) system can be used to qualitatively understand the phase chemistry of the noble-metal inclusions. The kinetics of the oxidation reaction are not particularly well understood, but they are of limited applicability to understanding the properties of used fuel. In contrast, it is important to determine the thermodynamic activity of molybdenum in noble-metal inclusions, so that analysis of their molybdenum content can be used as a probe of the local oxygen potential of the used fuel. (author)

Casting technology for manufacturing metal rods from simulated metallic spent fuels

Science.gov (United States)

Leeand, Y. S.; Lee, D. B.; Kim, C. K.; Shin, Y. J.; Lee, J. H.

2000-09-01

A uranium metal rod 13.5 mm in diameter and 1,150 mm long was produced from simulated metallic spent fuels with advanced casting equipment using the directional-solidification method. A vacuum casting furnace equipped with a four-zone heater to prevent surface oxidation and the formation of surface shrinkage holes was designed. By controlling the axial temperature gradient of the casting furnace, deformation by the surface shrinkage phenomena was diminished, and a sound rod was manufactured. The cooling behavior of the molten uranium was analyzed using the computer software package MAGMAsoft.

Low NOx Fuel Flexible Combustor Integration Project Overview

Science.gov (United States)

Walton, Joanne C.; Chang, Clarence T.; Lee, Chi-Ming; Kramer, Stephen

2015-01-01

The Integrated Technology Demonstration (ITD) 40A Low NOx Fuel Flexible Combustor Integration development is being conducted as part of the NASA Environmentally Responsible Aviation (ERA) Project. Phase 2 of this effort began in 2012 and will end in 2015. This document describes the ERA goals, how the fuel flexible combustor integration development fulfills the ERA combustor goals, and outlines the work to be conducted during project execution.

New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

International Nuclear Information System (INIS)

1963-01-01

One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1998-01-01

An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

Integrated fuel cell stack shunt current prevention arrangement

Energy Technology Data Exchange (ETDEWEB)

Roche, Robert P. (Cheshire, CT); Nowak, Michael P. (Bolton, CT)

1992-01-01

A fuel cell stack includes a plurality of fuel cells juxtaposed with one another in the stack and each including a pair of plate-shaped anode and cathode electrodes that face one another, and a quantity of liquid electrolyte present at least between the electrodes. A separator plate is interposed between each two successive electrodes of adjacent ones of the fuel cells and is unified therewith into an integral separator plate. Each integral separator plate is provided with a circumferentially complete barrier that prevents flow of shunt currents onto and on an outer peripheral surface of the separator plate. This barrier consists of electrolyte-nonwettable barrier members that are accommodated, prior to the formation of the integral separator plate, in corresponding edge recesses situated at the interfaces between the electrodes and the separator plate proper. Each barrier member extends over the entire length of the associated marginal portion and is flush with the outer periphery of the integral separator plate. This barrier also prevents cell-to-cell migration of any electrolyte that may be present at the outer periphery of the integral separator plate while the latter is incorporated in the fuel cell stack.

Integrity of spent CANDU fuel during and following dry storage

International Nuclear Information System (INIS)

Villagran, J.E.

2004-01-01

This report examines the issue of CANDU fuel integrity at the back end of the fuel cycle and outlines a program designed to provide assurance that used CANDU fuel will retain its integrity over an extended period. In specific terms, the program is intended to provide assurance that during and following extended dry storage the fuel will remain fit to undergo, without loss of integrity, the handling, packaging and transportation operations that might be necessary until it is placed in disposal containers. The first step in the development of the program was a review of the available technical information on phenomena relevant to fuel integrity. The major conclusions from that review were the following: Under normal storage conditions it is unlikely that the spent fuel will suffer significant degradation during a one-hundred year period and it should be possible to retrieve, repackage and transport the fuel as required, using methods and systems similar to those used today. However, to provide increased confidence regarding the above conclusion, investigations should be conducted in areas where there is higher uncertainty in the prediction of fuel condition and on some degradation processes to which the fuel appears to present higher vulnerability. The proposed program includes, among other tasks, irradiated fuel tests, analytical studies on the most relevant fuel degradation processes and the development of a long-term fuel verification program. (Author)

High temperature corrosion of metallic interconnects in solid oxide fuel cells

International Nuclear Information System (INIS)

Bastidas, D. M.

2006-01-01

Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC) instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects. However, metallic interconnects continue to be degraded despite the lowered temperature, and their corrosion products contribute to electrical degradation in the fuel cell. coatings of nickel, chromium, aluminium, zinc, manganese, yttrium or lanthanum between the interconnect and the electrodes reduce this degradation during operation. (Author) 66 refs

Water storage of liquid-metal fast-breeder-reactor fuel

International Nuclear Information System (INIS)

Meacham, S.A.

1982-01-01

The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry

Failed fuel identification techniques for liquid-metal cooled reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

1995-01-01

The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

Fuel Cell Development and Test Laboratory | Energy Systems Integration

Science.gov (United States)

Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

Recovery and utilization of valuable metals from spent nuclear fuel. 3: Mutual separation of valuable metals

International Nuclear Information System (INIS)

Kirishima, K.; Shibayama, H.; Nakahira, H.; Shimauchi, H.; Myochin, M.; Wada, Y.; Kawase, K.; Kishimoto, Y.

1993-01-01

In the project ''Recovery and Utilization of Valuable Metals from Spent Fuel,'' mutual separation process of valuable metals recovered from spent fuel has been studied by using the simulated solution contained Pb, Ru, Rh, Pd and Mo. Pd was separated successfully by DHS (di-hexyl sulfide) solvent extraction method, while Pb was recovered selectively from the raffinate by neutralization precipitation of other elements. On the other hand, Rh was roughly separated by washing the precipitate with alkaline solution, so that Rh was refined by chelate resin CS-346. Outline of the mutual separation process flow sheet has been established of the combination of these techniques. The experimental results and the process flow sheet of mutual separation of valuable metals are presented in this paper

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

Science.gov (United States)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Preliminary study or RSG-GAS reactor fuel element integrity

International Nuclear Information System (INIS)

Soejoedi, A.; Tarigan, A.; Sujalmo; Prayoga, S.; Suhadi

1996-01-01

After 8 years of operation, RSG-GAS was able to reach 15 cycles of reactor operation with 116 irradiated fuels, whereas 49 fuels were produced by NUKEM; and the other 67 were produced by PEBN-BATAN. At the 15 T h cycles, it have been used 40 standard fuels and 8 control fuels (Forty standard fuels and eight control fuels have been used in the 15 t h core cycles). Several activities have been performed in the reactor, to investigate the fuel integrity, among of them are: .fuel visual test with under water camera, which the results were recorder in the video cassette, primary water quality test during, reactor operation, fuel failure detector system examination and compared the PIE results in the Radiometallurgy Installation (RMI). The results showed that the fuel integrity, before and after irradiation, have still good performance and the fission products have not been released yet

Integrated data base for spent fuel and radwaste: inventories

International Nuclear Information System (INIS)

Notz, K.J.; Carter, W.L.; Kibbey, A.H.

1982-01-01

The Integrated Data Base (IDB) program provides and maintains current, integrated data on spent reactor fuel and radwaste, including historical data, current inventories, projected inventories, and material characteristics. The IDB program collects, organizes, integrates, and - where necessary - reconciles inventory and projection (I/P) and characteristics information to provide a coherent, self-consistent data base on spent fuel and radwaste

The miscibility and oxidation study of the simulated metallic spent fuel for the development of an advanced spent fuel management process

Energy Technology Data Exchange (ETDEWEB)

Shin, Y. J.; You, G. S.; Ju, J. S.; Lee, E. P.; Seo, H. S.; Ahn, S. B. [Korea Atomic Energy Research Institute, Taejeon (Korea)

1999-03-01

The simulated metallic spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the immiscibility of the some elements with metal uranium. 2 refs., 45 figs. (Author)

An integrated heavy metal emission inventory in Alsace and Baden-Wurtemberg

International Nuclear Information System (INIS)

Veaux, Ch.; Zundel, T.; Rentz, O.

1997-01-01

The emissions of the heavy metals As, Cd, Cr, Cu, Hg, Ni, Pb, Se, V, and Zn into the air, water and by-products, caused by stationary combustion plants in Baden-Wurtemberg (Germany) and Alsace (France) aRe assessed. The main emphasis of the study is on the development of an adequate, process specific tool allowing to determine multi-media heavy metal emissions by bottom ash or slag, filter ash, by-products of the flue gas desulfurization device (FGD by-products), and flue gas from the use of fossil fuels in Baden-Wurtemberg and Alsace. The heavy metal mass streams are analysed systematically with regard to fuel composition, boiler type, operation mode and flue gas treatment devices to define representative plants with regard to the partitioning of heavy metal streams. For these representative plants, partition factors, which characterise the distribution of heavy metals streams. The emission factors are derived from the partition factors and the heavy concentrations in the fuels used. Then, the emission factors are associated with data on emission sources referring to the configuration of combustion installations and activity data (annual consumption of fuel) to assess the heavy metal emissions in the study region. Heavy metal emissions through bottom ash, filter ash, FGD by-products and flue gas as well as heavy metal deposition in plants are assessed for the sectors 'public power plants', industrial combustion' and 'domestic heating' for the base year 1992. The results are available for Baden-Wurtemberg and Alsace for each heavy metal, sector, emission pathway, and combustion technique. (authors)

Research on plant of metal fuel fabrication using casting process

International Nuclear Information System (INIS)

Senda, Yasuhide; Mori, Yukihide

2003-12-01

This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

Corrosion of metal bipolar plates for PEM fuel cells: A review

Energy Technology Data Exchange (ETDEWEB)

Antunes, Renato A. [Engenharia de Materiais, Universidade Federal do ABC (UFABC), 09210-170 Santo Andre, SP (Brazil); Oliveira, Mara Cristina L.; Ett, Gerhard; Ett, Volkmar [Electrocell Ind. Com. Equip. Elet. LTDA, Centro de Inovacao, Empreendedorismo e Tecnologia (CIETEC), 05508-000 Sao Paulo, SP (Brazil)

2010-04-15

PEM fuel cells are of prime interest in transportation applications due to their relatively high efficiency and low pollutant emissions. Bipolar plates are the key components of these devices as they account for significant fractions of their weight and cost. Metallic materials have advantages over graphite-based ones because of their higher mechanical strength and better electrical conductivity. However, corrosion resistance is a major concern that remains to be solved as metals may develop oxide layers that increase electrical resistivity, thus lowering the fuel cell efficiency. This paper aims to present the main results found in recent literature about the corrosion performance of metallic bipolar plates. (author)

Influence of metallic based fuel additives on performance and exhaust emissions of diesel engine

Energy Technology Data Exchange (ETDEWEB)

Keskin, Ali [Tarsus Technical Education Faculty, Mersin University, 33500 Mersin (Turkey); Guerue, Metin, E-mail: mguru@gazi.edu.t [Engineering and Architectural Faculty, Gazi University, 06570 Maltepe, Ankara (Turkey); Altiparmak, Duran [Technical Education Faculty, Gazi University, 06500 Ankara (Turkey)

2011-01-15

In this experimental study, influence of the metallic-based additives on fuel consumption and exhaust emissions of diesel engine were investigated. The metallic-based additives were produced by synthesizing of resin acid (abietic acid) with MnO{sub 2} or MgO. These additives were doped into diesel fuel at the rate of 8 {mu}mol/l and 16 {mu}mol/l for preparing test fuels. Both additives improved the properties of diesel fuel such as viscosity, flash point, cloud point and pour point. The fuels with and without additives were tested in a direct injection diesel engine at full load condition. Maximum reduction of specific fuel consumption was recorded as 4.16%. CO emission and smoke opacity decreased by 16.35% and by 29.82%, respectively. NO{sub x} emission was measured higher and CO{sub 2} emission was not changed considerably with the metallic-based additives.

Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok

2015-01-01

The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature

Toward a sustainable energy supply with reduced environmental burden. Development of metal fuel fast reactor cycle

International Nuclear Information System (INIS)

Koyama, Tadafumi; Kobayashi, Hiroaki; Kinoshita, Kensuke

2009-01-01

CRIEPI has been studying the metal fuel fast reactor cycle as an outstanding alternative for the future energy sources. In this paper, development of the metal fuel cycle is reviewed in the view point of technological feasibility and material balance. Preliminary estimation of reduction of the waste burden due to introduction of the metal fuel cycle technology is also reported. (author)

Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

The performance of 3500 MWth homogeneous and heterogeneous metal fueled core designs

International Nuclear Information System (INIS)

Turski, R.; Yang, Shi-tien.

1987-11-01

Performance parameters are calculated for a representative 3500 MWth homogeneous and a heterogeneous metal fueled reactor design. The equilibrium cycle neutronic characteristics, safety coefficients, control system requirements, and control rod worths are evaluated. The thermal-hydraulic characteristics for both configurations are also compared. The heavy metal fuel loading requirements and neutronic performance characteristics are also evaluated for the uranium startup option. 14 refs., 14 figs., 20 tabs

Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

International Nuclear Information System (INIS)

Zhang, Youpeng; Wallenius, Janne

2014-01-01

Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs

Refinery Integration of By-Products from Coal-Derived Jet Fuels

Energy Technology Data Exchange (ETDEWEB)

Caroline E. Burgess Clifford; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

2006-05-17

This report summarizes the accomplishments toward project goals during the first six months of the third year of the project to assess the properties and performance of coal based products. These products are in the gasoline, diesel and fuel oil range and result from coal based jet fuel production from an Air Force funded program. Specific areas of progress include generation of coal based material that has been fractionated into the desired refinery cuts, acquisition and installation of a research gasoline engine, and modification of diesel engines for use in evaluating diesel produced in the project. Characterization of the gasoline fuel indicates a dominance of single ring alkylcycloalkanes that have a low octane rating; however, blends containing these compounds do not have a negative effect upon gasoline when blended in refinery gasoline streams. Characterization of the diesel fuel indicates a dominance of 3-ring aromatics that have a low cetane value; however, these compounds do not have a negative effect upon diesel when blended in refinery diesel streams. The desulfurization of sulfur containing components of coal and petroleum is being studied so that effective conversion of blended coal and petroleum streams can be efficiently converted to useful refinery products. Equipment is now in place to begin fuel oil evaluations to assess the quality of coal based fuel oil. Combustion and characterization of fuel oil indicates that the fuel is somewhere in between a No. 4 and a No. 6 fuel oil. Emission testing indicates the fuel burns similarly to these two fuels, but trace metals for the coal-based material are different than petroleum-based fuel oils. Co-coking studies using cleaned coal are highly reproducible in the pilot-scale delayed coker. Evaluation of the coke by Alcoa, Inc. indicated that while the coke produced is of very good quality, the metals content of the carbon is still high in iron and silica. Coke is being evaluated for other possible uses

Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

International Nuclear Information System (INIS)

Sugino, Kazuteru; Mizuno, Tomoyasu

2003-06-01

Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

Fuels planning: science synthesis and integration; fact sheet: The Fuels Synthesis Project overview

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

The geographic focus of the "Fuels Planning: Science Synthesis and Integration" project #known as the Fuels Synthesis Project# is on the dry forests of the Western United States. Target audiences include fuels management specialists, resource specialists, National Environmental Policy Act #NEPA# planning team leaders, line officers in the USDA Forest Service...

Application of Ceramic Bond Coating for Reusable Melting Crucible of Metallic Fuel Slugs

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki-Hwan; Song, Hoon; Ko, Young-Mo; Park, Jeong-Yong; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hong, Ki-Won [Chungnam National University, Daejeon (Korea, Republic of)

2016-10-15

Metal fuel slugs of the driver fuel assembly have been fabricated by injection casting of the fuel alloys under a vacuum state or an inert atmosphere. Traditionally, metal fuel such as a U-Zr alloy system for SFR has been melted in slurry-coated graphite crucibles and cast in slurry-coated quartz tube molds to prevent melt/material interactions. Reactive coatings and porous coatings can be a source of melt contaminations, and fuel losses, respectively. Ceramic Y{sub 2}O{sub 3}, TiC, and TaC coating materials showed no penetration in the protective layer after a melt dipping test. However, the ceramic coating materials showed separations in the coating interface between the substrate and coating layer, or between the coating layer and fuel melt after the dipping test. All plasma-spray coated methods maintained a sound coating state after a dipping test with U-10wt.%Zr melt. A single coating Y{sub 2}O{sub 3}(150) layer and double coating layer of TaC(50)-Y{sub 2}O{sub 3}(100), showed a sound state or little penetration in the protective layer after a dipping test with U-10wt.%Zr-5wt.%RE melt. Injection casting experiments of U-10wt.%Zr and U-10wt.%Zr-5wt.%RE fuel slugs have been performed to investigate the feasibility of a reusable crucible of the metal fuel slugs. U–10wt.%Zr and U–10wt.%Zr–5wt.%RE fuel slugs have been soundly fabricated without significant interactions of the graphite crucibles. Thus, the ceramic plasma-spray coatings are thought to be promising candidate coating methods for a reusable graphite crucible to fabricate metal fuel slugs.

Modeling of constituent redistribution in U-Pu-Zr metallic fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2006-12-01

A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

Conceptual assessment and thermal hydraulic analysis of MVDS system for the dry storage of reduced metal fuel

International Nuclear Information System (INIS)

Lee, J. C.; Bang, K. S.; Shin, H. S.; Joo, J. S.; Su, K. S.; Kim, H. D.

2003-01-01

Conceptual assessment and thermal hydraulic analysis of MVDS storage system have been carried out for application of reduced metal fuel. The storage concept was established considering the optimum weight, storage volume and thermal efficiency. The capacity of MVDS system for loading the reduced metal fuel has four times as compared with existing PWR fuel storage system. In the results of thermal analysis, the maximum temperature of metal fuel was estimated to be 110 .deg. C which is lower than the allowable value under normal operation condition. Therefore, it is shown that the MVDS system can feasibly accomodate the reduced metal fuel in aspect of thermal safety

Metallic Fuels Handbook

Energy Technology Data Exchange (ETDEWEB)

Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burkes, Douglas E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cole, James I. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fielding, Randall S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frank, Steven M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hartmann, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hyde, Timothy A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, J. Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddison, Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mariani, Robert D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Middlemas, Scott C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Holleran, Thomas P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sencer, Bulent H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Squires, Leah N. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-08-07

This is not a typical External Report--It is a Handbook. No Abstract is involved. This includes both Parts 1 and 2. The Metallic Fuels Handbook summarizes currently available information about phases and phase diagrams, heat capacity, thermal expansion, and thermal conductivity of elements and alloys in the U-Pu-Zr-Np-Am-La-Ce-Pr-Nd system. Although many sections are reviews and updates of material in previous versions of the Handbook [1, 2], this revision is the first to include alloys with four or more elements. In addition to presenting information about materials properties, the handbook attempts to provide information about how well each property is known and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data.

The integrity of CANDU fuel during load following

International Nuclear Information System (INIS)

Tayal, M.; Manzer, A.M.; Sejnoha, R.; Hains, A.J.

1989-08-01

This paper summarizes data and analyses of integrity and of physics of CANDU fuel during load following. Measurements of irradiated fuel show that power cycles do not enhance release of fission gas. Data from research reactors show that the power cycles cause cyclic strains in the sheath. Finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath. The stresses and the strains are well into the plastic range. The cyclic loads 'use up' some fraction of the sheath's resistance to environmentally-assisted cracking (EAC), depending on the details of the fuel design and of then power cycles. The balance of the sheath's resistance to EAC continues to be available to counteract static loads. Thousands of fuel bundles have experienced many power cycles in research and in commercial reactors. Overall integrity of fuel bundles is well over 99%. Thus, CANDU fuel continues to show good performance in both base-load and load-following reactors

Fuel upgrading and reforming with metal organic framework

KAUST Repository

Eddaoudi, Mohamed

2016-03-31

Systems and methods for separating hydrocarbons on an internal combustion powered vehicle via one or more metal organic frameworks are disclosed. Systems and methods can further include utilizing separated hydrocarbons and exhaust to generate hydrogen gas for use as fuel. In one aspect, a method for separating hydrocarbons can include contacting a first component containing a first metal organic framework with a flow of hydrocarbons and separating hydrocarbons by size. In certain embodiments, the hydrocarbons can include alkanes.

Progress on the Application of Metallic Fuels for Actinide Transmutation

International Nuclear Information System (INIS)

Kennedy, J. Rory; Fielding, Randall; Janney, Dawn; Mariani, Robert; Teague, Melissa; Egeland, Gerald

2015-01-01

Full text of publication follows: Idaho National Laboratory (INL) is developing actinide bearing alloy metallic fuels intended for effecting the transmutation of long-lived isotopes in fast reactor application as part of a partitioning and transmutation strategy. This presentation will report on progress in three areas of this effort: demonstration of the fabrication of fuels under remote (hot cell) conditions directly coupled to the product from the Pyro-processing of spent fuel as part of the Joint Fuel Cycle Studies (JFCS) collaboration with the Korean Atomic Energy Research Institute (KAERI); the chemical sequestration of lanthanide fission products to mitigate fuel-cladding-chemical-interaction (FCCI); and transmission electron microscopy (TEM) and atom probe tomography (APT) studies on the as-cast microstructure of the metallic fuel alloy. For the JFCS efforts, we report on the implementation of the Glove-box Advanced Casting System (GACS) as a prototype casting furnace for eventual installation into the INL Hot Fuel Examination Facility (HFEF) where the recycled fuel will be cast. Results from optimising process parameters with respect to fuel characteristics, americium volatility, materials interaction, and lanthanide fission product carry over distribution will be discussed. With respect to the lanthanide carry over from the Pyro-processing product, encouraging studies on concepts to chemically sequester the FCCI promoting lanthanides within the fuel matrix thus inhibiting migration and interaction with the cladding will be presented. Finally, in relation to advanced modelling and simulation efforts, detailed investigations and interpretation on the nano-scale as cast microstructure of possible recycle fuel composition containing U, Pu, Am, Np as well as carry-over lanthanide species will be discussed. These studies are important for establishing the initial conditions from which advanced physics based fuel performance codes will run. (authors)

On-line fuel and control rod integrity management in BWRs

International Nuclear Information System (INIS)

Larsson, Irina; Sihver, Lembit

2011-01-01

Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of a nuclear power plant. An accurate and prompt way to monitor fuel integrity in a reactor core during reactor operation is by using on-line measurements of the gamma emitting noble gas activities in the off-gas system. The integrity of control rods can be efficiently followed by on-line measurements of the helium (He) concentration in the off-gases. This method also gives information about fuel rod failures since He is used as a fill gas in the fuel rods. To survey fuel and control rod integrity during reactor operation, a system consisting of combined gamma and He on-line measurements in the off-gases should be used. Such a system can detect and follow the behavior of fuel and control rod failures. In addition, it can separate fuel failures from control rod failures since fuel rods contain both He and gamma emitting noble gases, while control rods only contain He. Moreover, the system is able to distinguish primary fuel failures from degradation of already existing ones. In this paper we present a combined system for on-line measurements of He and gamma emitting noble gases in the reactor off-gas system and measuring experiences from different BWRs. (author)

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

Fabrication of preliminary fuel rods for SFR

International Nuclear Information System (INIS)

Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

2012-01-01

Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

Integral Fast Reactor concept

Energy Technology Data Exchange (ETDEWEB)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

Integral Fast Reactor concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path

Development of code SFINEL (Spent fuel integrity evaluator)

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

1999-01-01

SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1991-01-01

In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1990-11-01

The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

33 CFR 183.550 - Fuel tanks: Installation.

Science.gov (United States)

2010-07-01

...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.550 Fuel tanks: Installation. (a) Each fuel tank must not be integral with any boat structure or mounted on an engine. (b) Each... the top surface of each metallic fuel tank when the boat is in its static floating position. (e) Each...

Fast reactors with axial arrangement of oxide and metal fuels in the core

International Nuclear Information System (INIS)

Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

1980-01-01

Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels

International Nuclear Information System (INIS)

Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.

1996-01-01

The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)

Consequences of metallic fuel-cladding liquid phase attack during over-temperature transient on fuel element lifetime

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Seidel, B.R.

1990-01-01

Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs

Fuel containing vessel for transporting nuclear fuel

International Nuclear Information System (INIS)

Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

1996-01-01

A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

2010-01-31

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

Inherent safe fast breeder reactors and actinide burners, metallic fuel

International Nuclear Information System (INIS)

Dorner, S.; Schumacher, G.

1991-04-01

Nuclear power without breeder strategy uses the possibilities for the energy supply only to a small extend compared to the possibilities of fast breeder reactors, which offer an energy supply for thousands of years. Moreover, a fast neutron device offers the opportunity to run an actinide-burner that could improve the situation of waste management. Within this concept metallic fuel could play a key role. The present report shows some important aspects of the concept like the pyrometallic reprocessing, the behaviour of metallic fuel during a core meltdown accident and others. The report should contribute to the discussion of these problems and initialize further work

Development of metal fuel and study of construction materials (I-IV), Part II

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

The studies were devoted to problems related to application of metal uranium as fuel in heavy water reactors. Influence of thermal treatment on material texture and recrystallization of cast uranium was investigated. Structural changes of uranium alloys with molybdenum and niobium were tested during different heat treatments. A review of the possibilities for using metal uranium fuel in heavy water reactors is included

Coal Integrated Gasification Fuel Cell System Study

Energy Technology Data Exchange (ETDEWEB)

Gregory Wotzak; Chellappa Balan; Faress Rahman; Nguyen Minh

2003-08-01

The pre-baseline configuration for an Integrated Gasification Fuel Cell (IGFC) system has been developed. This case uses current gasification, clean-up, gas turbine, and bottoming cycle technologies together with projected large planar Solid Oxide Fuel Cell (SOFC) technology. This pre-baseline case will be used as a basis for identifying the critical factors impacting system performance and the major technical challenges in implementing such systems. Top-level system requirements were used as the criteria to evaluate and down select alternative sub-systems. The top choice subsystems were subsequently integrated to form the pre-baseline case. The down-selected pre-baseline case includes a British Gas Lurgi (BGL) gasification and cleanup sub-system integrated with a GE Power Systems 6FA+e gas turbine and the Hybrid Power Generation Systems planar Solid Oxide Fuel Cell (SOFC) sub-system. The overall efficiency of this system is estimated to be 43.0%. The system efficiency of the pre-baseline system provides a benchmark level for further optimization efforts in this program.

Physics studies of weapons plutonium disposition in the Integral Fast Reactor closed fuel cycle

International Nuclear Information System (INIS)

Hill, R.N.; Wade, D.C.; Liaw, J.R.; Fujita, E.K.

1995-01-01

The core performance impact of weapons plutonium introduction into the Integral Fast Reactor (IFR) closed fuel cycle is investigated by comparing three disposition scenarios: a power production mode, a moderate destruction mode, and a maximum destruction mode, all at a constant heat rating of 840 MW(thermal). For each scenario, two fuel cycle models are evaluated: cores using weapons material as the sole source of transuranics in a once-through mode and recycle cores using weapons material only as required for a makeup feed. In addition, the impact of alternative feeds (recycled light water reactor or liquid-metal reactor transuranics) on burner core performance is assessed. Calculated results include mass flows, detailed isotopic distributions, neutronic performance characteristics, and reactivity feedback coefficients. In general, it is shown that weapons plutonium does not have an adverse effect on IFR core performance characteristics; also, favorable performance can be maintained for a wide variety of feed materials and fuel cycle strategies

Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

International Nuclear Information System (INIS)

Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

1991-01-01

An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

Possibility of using metal uranium fuel in heavy water reactors

International Nuclear Information System (INIS)

Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

1965-01-01

The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

In situ metal ion contamination and the effects on proton exchange membrane fuel cell performance

Science.gov (United States)

Sulek, Mark; Adams, Jim; Kaberline, Steve; Ricketts, Mark; Waldecker, James R.

Automotive fuel cell technology has made considerable progress, and hydrogen fuel cell vehicles are regarded as a possible long-term solution to reduce carbon dioxide emissions, reduce fossil fuel dependency and increase energy efficiency. Even though great strides have been made, durability is still an issue. One key challenge is controlling MEA contamination. Metal ion contamination within the membrane and the effects on fuel cell performance were investigated. Given the possible benefits of using stainless steel or aluminum for balance-of-plant components or bipolar plates, cations of Al, Fe, Ni and Cr were studied. Membranes were immersed in metal sulfide solutions of varying concentration and then assembled into fuel cell MEAs tested in situ. The ranking of the four transition metals tested in terms of the greatest reduction in fuel cell performance was: Al 3+ ≫ Fe 2+ > Ni 2+, Cr 3+. For iron-contaminated membranes, no change in cell performance was detected until the membrane conductivity loss was greater than approximately 15%.

The US Liquid-Metal Reactor Program - overview and status

International Nuclear Information System (INIS)

Quinn, J.E.; Gyorey, G.L.; Salerno, L.N.

1992-01-01

The US Advanced Liquid-Metal Reactor (ALMR) Program has three major elements being developed in an integrated fashion to produce a system meeting the US long-term nuclear energy needs. Reactor design, one of those elements, is the focus of this paper. The other two elements, the integral fast reactor metal-fuel cycle and the light water reactor (LWR) spent-fuel actinide recycle, will be addressed in companion papers. The ALMR is adaptable to multiple missions with few modifications such as the core arrangements. The missions identified to date are (a) the extension of the existing uranium resources through breeding and highly efficient uranium utilization, (b) the recycle and utilization of the long-life actinides in LWR spent fuel as fissile material for the ALMR, and (c) the conversion of excess weapons fissil material into electricity. In addition to these missions, the reactor design is adaptable to either the metal-fuel cycle or the oxide fuel cycle

The Integral Fast Reactor concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes the key features and potential advantages of the IFR concept, its technology development status, fuel cycle economics potential, and its future development path

Description of reactor fuel breeding with three integral concepts

International Nuclear Information System (INIS)

Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

1979-01-01

The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

IFR fuel cycle--pyroprocess development

International Nuclear Information System (INIS)

Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

1992-01-01

The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

Evolution of the liquid metal reactor

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

This paper reports on the integral fat reactor (IFR) concept. A key feature of the IFR concept is the metallic fuel, the original choice in liquid metal reactor development. An IFR development program is detailed by the authors

Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

International Nuclear Information System (INIS)

Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

1993-10-01

An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

International Nuclear Information System (INIS)

Herrmann, S.D.; Li, S.X.

2010-01-01

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl - 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

2000-01-01

Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

PYRO, a system for modeling fuel reprocessing

International Nuclear Information System (INIS)

Ackerman, J.P.

1989-01-01

Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions

Literature survey on metal waste form for metallic waste from electrorefiners for the electrometallurgical treatment of spent metallic fuels

International Nuclear Information System (INIS)

Nishimura, Tomohiro

2003-01-01

This report summarizes the recent results of the metal waste form development activities at the Argonne National Laboratory in the USA for high-level radioactive metallic waste (stainless-steel (SS) cladding hulls, zirconium (Zr), noble-metal fission products (NMFPs), etc.) from electrorefiners for the electrometallurgical treatment of spent metallic fuels. Their main results are as follows: (1) SS- 15 wt.% Zr- ∼4 wt.% NMFPs alloy was selected as the metal waste form, (2) metallurgical data, properties, long-term corrosion data, etc. of the alloy have been collected, (3) 10-kg ingots have been produced in hot tests and a 60-kg production machine is under development. The following research should be made to show the feasibility of the metal waste form in Japan: (1) degradation assessment of the metal waste form in Japanese geological repository environments, and (2) clarification of the maximum allowable contents of NMFPs. (author)

Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.

1985-01-01

A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

The possibility of prediction of the lifetime of metallic nuclear fuel elements in a radiation field of thermal nuclear reactors

International Nuclear Information System (INIS)

Livne, Z.; Ramon, P.

1979-01-01

An attempt is made to clarify the possible causes of failure of irradiated nuclear fuel cartridges, in order to determine the parameters which govern the lifetime of the fuel and a way to predict it. Measurements of mechanical properties of irradiated uranium metal and cladding, can serve as a basis for failure prediction. Testing irradiated fuel elements by bending till fracture enables to evaluate the integral character of the fuel element, along the cross-section, taking into account the difference in brittleness of several zones. It is likely that the bending test, which indicates the behaviour of a stress-strain function, is a faster and more reliable way to determine the mechanical properties of irradiated nuclear fuel. Since the stresses applied to the cladding during irradiation are locally hydrostatic, its postirradiation blow-up provide information on strength and elasticity variations of the irradiated cladding material. (B.G.)

Development of fast reactor metal fuels containing minor actinides

International Nuclear Information System (INIS)

Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

2011-01-01

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

Carbon composites with metal nanoparticles for Alcohol fuel cells

Science.gov (United States)

Ventrapragada, Lakshman; Siddhardha, R. S.; Podilla, Ramakrishna; Muthukumar, V. S.; Creager, Stephen; Rao, A. M.; Ramamurthy, Sai Sathish

2015-03-01

Graphene due to its high surface area and superior conductivity has attracted wide attention from both industrial and scientific communities. We chose graphene as a substrate for metal nanoparticle deposition for fuel cell applications. There are many chemical routes for fabrication of metal-graphene composites, but they have an inherent disadvantage of low performance due to the usage of surfactants, that adsorb on their surface. Here we present a design for one pot synthesis of gold nanoparticles and simultaneous deposition on graphene with laser ablation of gold strip and functionalized graphene. In this process there are two natural advantages, the nanoparticles are synthesized without any surfactants, therefore they are pristine and subsequent impregnation on graphene is linker free. These materials are well characterized with electron microscopy to find their morphology and spectroscopic techniques like Raman, UV-Vis. for functionality. This gold nanoparticle decorated graphene composite has been tested for its electrocatalytic oxidation of alcohols for alkaline fuel cell applications. An electrode made of this composite showed good stability for more than 200 cycles of operation and reported a low onset potential of 100 mV more negative, an important factor for direct ethanol fuel cells.

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

2009-03-01

The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON

International Nuclear Information System (INIS)

Galloway, J.; Unal, C.; Carlson, N.; Porter, D.; Hayes, S.

2015-01-01

Highlights: • An improved constituent distribution formulation in metallic nuclear fuels. • The new algorithm is implemented into the advanced fuel performance framework BISON. • Experimental Breeder Reactor-II data, T179, DP16, T459 are reanalyzed. • Phase dependent diffusion coefficients are improved. • Most influential phase is gamma, followed by alpha and thirdly the beta phase. - Abstract: An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor-II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin finite element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U–Pu–Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all

14 CFR 23.967 - Fuel tank installation.

Science.gov (United States)

2010-01-01

... the engine compartment may act as the wall of an integral tank. (d) Each fuel tank must be isolated... loads without permanent deformation or failure under the conditions of §§ 23.365 and 23.843 of this part. A bladder-type fuel cell, if used, must have a retaining shell at least equivalent to a metal fuel...

Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

Possibilities of using metal uranium fuel in heavy water reactors

International Nuclear Information System (INIS)

Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

1965-11-01

There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

Establishment of Experimental Apparatus and Mechanical Test for SFR Metallic Fuel

International Nuclear Information System (INIS)

Kim, Sun Ki; Lee, Chong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Yoon Myung; Lee, Chan Bock

2010-12-01

U-Zr binary alloys and U-Zr-Ce ternary alloys as SFR surrogate metallic fuels were fabricated by a casting process. Tensile tests were performed to evaluate the mechanical properties of the fuels. As a results, the mechanical properties such as yield strength, ultimate tensile strength, and elongation were measured. In this report, these experimental results are presented

Demonstration of pyrometallurgical processing for metal fuel and HLW

International Nuclear Information System (INIS)

Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

2001-01-01

CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

Indigenous development of system integration for proton exchange membrane fuel cell operation

International Nuclear Information System (INIS)

Hussain, S.; Arshad, M.; Anjum, A.R.

2011-01-01

System integration was developed for fuel cell to control various parameters including voltage, current, power, temperature, pressure of gas (H/sub 2/), humidification, etc. The compact software has also been developed for monitoring different parameters of fuel cell system. System integrated was installed on fuel cell stack to manipulate these parameters. The compact software has been linked with the integrated system for visual monitoring of different parameters of fuel cell system during operation on PC. The installation of software and integrated system on fuel cell stack is the key achievement for the safe operation of fuel cell stack and for the provision of requisite power to any electric device for optimum performance. The compact software was developed for micro controller in KIEL. Control card and driver card are controlled by software-driven micro controller. A communication protocol was designed and developed. PC software has been developed to control and watch the values of all parameters of fuel cell such as voltage, current, power, temperature, pressure of hydrogen, pressure of oxygen, operational times and performance of the system on computer screen. (author)

A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

International Nuclear Information System (INIS)

Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua

2017-01-01

In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design

A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

Energy Technology Data Exchange (ETDEWEB)

Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang (China)

2017-06-15

In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

International Nuclear Information System (INIS)

You, WuSeung; Hong, Ser Gi

2014-01-01

The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

Integration of fuel cells into residential buildings

International Nuclear Information System (INIS)

Bell, J.M.; Entchev, E.; Gusdorf, J.; Szadkowski, F.; Swinton, M.; Kalbfleisch, W.; Marchand, R.

2004-01-01

Integration of small combined heat and power systems (CHP) into residential buildings is challenging as the loads are small, the load diversity is limited and there are a number of unresolved issues concerning sizing, control, peak loads, emergency operation, grid connection and export, etc. Natural Resources Canada has undertaken an initiative to investigate and develop techniques for the integration of small CHP systems into residential buildings using a highly instrumented house modified to allow quick installation and thorough monitoring of CHP integration techniques as well determining the performance of the CHP systems themselves when operating in a house. The first CHP system installed was a Stirling engine residential CHP system. It was used to examine the completeness of the CHP modifications to the house, to evaluate various building integration techniques and to measure the performance of the CHP system itself. The testing demonstrated the modified house to be an excellent facility for the development of CHP building integration techniques and the testing of residential CHP systems. The Stirling engine CHP system was found to operate well and produce meaningful input to the house. A second system (residential fuel cell) is presently being installed and building integration techniques and the performance of the fuel cell will be tested over the coming year. (author)

Integral fast reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Integral Fast Reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Shin, Young Joon; Cho, S. H.; You, G. S.

2001-04-01

Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

International Nuclear Information System (INIS)

Pierce, V.H.

2005-01-01

1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

Integration of the military and civilian nuclear fuel cycles in Russia

International Nuclear Information System (INIS)

Bukharin, O.

1994-01-01

This paper describes the close integration of the civil and military nuclear fuel cycles in Russia. Individual processing facilities, as well as the flow of nuclear material, are described as they existed in the 1980s and as they exist today. The end of the Cold War and the breakup of the Soviet Union weakened the ties between the two nuclear fuel cycles, but did not separate them. Separation of the military and civilian nuclear fuel cycles would facilitate Russia's integration into the world's nuclear fuel cycle and its participation in international non-proliferation regimes

Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

International Nuclear Information System (INIS)

Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

International Nuclear Information System (INIS)

Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

2003-01-01

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

Application of fuel cells with heat recovery for integrated utility systems

Science.gov (United States)

Shields, V.; King, J. M., Jr.

1975-01-01

This paper presents the results of a study of fuel cell powerplants with heat recovery for use in an integrated utility system. Such a design provides for a low pollution, noise-free, highly efficient integrated utility. Use of the waste heat from the fuel cell powerplant in an integrated utility system for the village center complex of a new community results in a reduction in resource consumption of 42 percent compared to conventional methods. In addition, the system has the potential of operating on fuels produced from waste materials (pyrolysis and digester gases); this would provide further reduction in energy consumption.

Integrated Fuel-Coolant Interaction (IFCI 6.0) code

International Nuclear Information System (INIS)

Davis, F.J.; Young, M.F.

1994-04-01

The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

Implementation, verification, and validation of the FPIN2 metal fuel pin mechanics model in the SASSYS/SAS4A LMR transient analysis codes

International Nuclear Information System (INIS)

Sofu, T.; Kramer, J.M.

1994-01-01

The metal fuel version of the FPIN2 code which provides a validated pin mechanics model is coupled with SASSYS/SAS4A Version 3.0 for single pin calculations. In this implementation, SASSY/SAS4A provides pin temperatures, and FPIN2 performs analysis of pin deformation and predicts the time and location of cladding failure. FPIN2 results are also used for the estimates of axial expansion of fuel and associated reactivity effects. The revalidation of the integrated SAS-FPIN2 code system is performed using TREAT tests

Calculating failure probabilities for TRISO-coated fuel particles using an integral formulation

International Nuclear Information System (INIS)

Miller, Gregory K.; Maki, John T.; Knudson, Darrell L.; Petti, David A.

2010-01-01

The fundamental design for a gas-cooled reactor relies on the safe behavior of the coated particle fuel. The coating layers surrounding the fuel kernels in these spherical particles, termed the TRISO coating, act as a pressure vessel that retains fission products. The quality of the fuel is reflected in the number of particle failures that occur during reactor operation, where failed particles become a source for fission products that can then diffuse through the fuel element. The failure probability for any batch of particles, which has traditionally been calculated using the Monte Carlo method, depends on statistical variations in design parameters and on variations in the strengths of coating layers among particles in the batch. An alternative approach to calculating failure probabilities is developed herein that uses direct numerical integration of a failure probability integral. Because this is a multiple integral where the statistically varying parameters become integration variables, a fast numerical integration approach is also developed. In sample cases analyzed involving multiple failure mechanisms, results from the integration methods agree closely with Monte Carlo results. Additionally, the fast integration approach, particularly, is shown to significantly improve efficiency of failure probability calculations. These integration methods have been implemented in the PARFUME fuel performance code along with the Monte Carlo method, where each serves to verify accuracy of the others.

Refinery Integration of By-Products from Coal-Derived Jet Fuels

Energy Technology Data Exchange (ETDEWEB)

Caroline Clifford; Andre Boehman; Chunshan Song; Bruce Miller; Gareth Mitchell

2008-03-31

The final report summarizes the accomplishments toward project goals during length of the project. The goal of this project was to integrate coal into a refinery in order to produce coal-based jet fuel, with the major goal to examine the products other than jet fuel. These products are in the gasoline, diesel and fuel oil range and result from coal-based jet fuel production from an Air Force funded program. The main goal of Task 1 was the production of coal-based jet fuel and other products that would need to be utilized in other fuels or for non-fuel sources, using known refining technology. The gasoline, diesel fuel, and fuel oil were tested in other aspects of the project. Light cycle oil (LCO) and refined chemical oil (RCO) were blended, hydrotreated to removed sulfur, and hydrogenated, then fractionated in the original production of jet fuel. Two main approaches, taken during the project period, varied where the fractionation took place, in order to preserve the life of catalysts used, which includes (1) fractionation of the hydrotreated blend to remove sulfur and nitrogen, followed by a hydrogenation step of the lighter fraction, and (2) fractionation of the LCO and RCO before any hydrotreatment. Task 2 involved assessment of the impact of refinery integration of JP-900 production on gasoline and diesel fuel. Fuel properties, ignition characteristics and engine combustion of model fuels and fuel samples from pilot-scale production runs were characterized. The model fuels used to represent the coal-based fuel streams were blended into full-boiling range fuels to simulate the mixing of fuel streams within the refinery to create potential 'finished' fuels. The representative compounds of the coal-based gasoline were cyclohexane and methyl cyclohexane, and for the coal-base diesel fuel they were fluorine and phenanthrene. Both the octane number (ON) of the coal-based gasoline and the cetane number (CN) of the coal-based diesel were low, relative to

Coal Integrated Gasification Fuel Cell System Study

Energy Technology Data Exchange (ETDEWEB)

Chellappa Balan; Debashis Dey; Sukru-Alper Eker; Max Peter; Pavel Sokolov; Greg Wotzak

2004-01-31

This study analyzes the performance and economics of power generation systems based on Solid Oxide Fuel Cell (SOFC) technology and fueled by gasified coal. System concepts that integrate a coal gasifier with a SOFC, a gas turbine, and a steam turbine were developed and analyzed for plant sizes in excess of 200 MW. Two alternative integration configurations were selected with projected system efficiency of over 53% on a HHV basis, or about 10 percentage points higher than that of the state-of-the-art Integrated Gasification Combined Cycle (IGCC) systems. The initial cost of both selected configurations was found to be comparable with the IGCC system costs at approximately $1700/kW. An absorption-based CO2 isolation scheme was developed, and its penalty on the system performance and cost was estimated to be less approximately 2.7% and $370/kW. Technology gaps and required engineering development efforts were identified and evaluated.

Expected savings from the relaxation of the limits on some fuel parameters for an HTR tubular interacting fuel element of fixed geometry

Energy Technology Data Exchange (ETDEWEB)

Graziani, [EURATOM, Ispra (Italy); Trauwaert, [Belgonucleaire (Belgium)

1971-06-16

This note presents some preliminary results of a recent work performed jointly by EURATOM (CCR Ispra) and BELGONUCLEAIRE. The analysis of the results is still in progress and a more comprehensive presentation will be made later. The purpose of the study was the evaluation of the savings in fuel cycle cost which could result from the relaxation of the limits on some fuels parameters - heavy metal loading, burnup and age factor - while keeping unchanged the geometry of the fuel element and that of the reactor. Within the range considered and the economic assumptions used, the most obvious saving which can be obtained from the increase of one of the limits seems that made possible by the increase of heavy metal loading. This saving is 2% in fuel cycle cost without any change in resonance integral and 4.5% with a reduction of 10% in resonance integral. An increase of the admissible burnup is also very rewarding, provided the resonance integral can be increased in order to keep the age factor down to 1.085.

Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

International Nuclear Information System (INIS)

Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

2005-01-01

The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Synthesis of uranium metal using laser-initiated reduction of uranium tetrafluoride by calcium metal

International Nuclear Information System (INIS)

West, M.H.; Martinez, M.M.; Nielsen, J.B.; Court, D.C.; Appert, Q.D.

1995-09-01

Uranium metal has numerous uses in conventional weapons (armor penetrators) and nuclear weapons. It also has application to nuclear reactor designs utilizing metallic fuels--for example, the former Integral Fast Reactor program at Argonne National Laboratory. Uranium metal also has promise as a material of construction for spent-nuclear-fuel storage casks. A new avenue for the production of uranium metal is presented that offers several advantages over existing technology. A carbon dioxide (CO 2 ) laser is used to initiate the reaction between uranium tetrafluoride (UF 4 ) and calcium metal. The new method does not require induction heating of a closed system (a pressure vessel) nor does it utilize iodine (I 2 ) as a chemical booster. The results of five reductions of UF 4 , spanning 100 to 200 g of uranium, are evaluated, and suggestions are made for future work in this area

Influence of oxygen-metal ratio on mixed-oxide fuel performance

International Nuclear Information System (INIS)

Lawrence, L.A.; Leggett, R.D.

1979-04-01

The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

Tubular solid oxide fuel cells with porous metal supports and ceramic interconnections

Science.gov (United States)

Huang, Kevin [Export, PA; Ruka, Roswell J [Pittsburgh, PA

2012-05-08

An intermediate temperature solid oxide fuel cell structure capable of operating at from 600.degree. C. to 800.degree. C. having a very thin porous hollow elongated metallic support tube having a thickness from 0.10 mm to 1.0 mm, preferably 0.10 mm to 0.35 mm, a porosity of from 25 vol. % to 50 vol. % and a tensile strength from 700 GPa to 900 GPa, which metallic tube supports a reduced thickness air electrode having a thickness from 0.010 mm to 0.2 mm, a solid oxide electrolyte, a cermet fuel electrode, a ceramic interconnection and an electrically conductive cell to cell contact layer.

Defining a metal-based waste form for IFR pyroprocessing wastes

International Nuclear Information System (INIS)

McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.

1994-01-01

Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts

Small-scale irradiated fuel electrorefining

International Nuclear Information System (INIS)

Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

1993-01-01

In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

Integral fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

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

A low cost liquid metal reactor design

International Nuclear Information System (INIS)

Arnold, W.H.; Anderson, C.A.; Mangus, J.D.

1984-01-01

A new, compact Liquid Metal Reactor (LMR) plant arrangement designed by Westinghouse, featuring factory-fabricated modules and an integrated fuel cycle facility, has made it possible to project a commercially competitive LMR plant for the near future. This innovative liquid metal-cooled plant design will allow a combination of capital, fuel, operation and maintenance costs that could be lower than today's fossil-fueled or light water reactor plant costs, and incorporate features which enhance public safety even beyond current high standards. Following early core loadings, the plant feeds only on depleted uranium. No shipment of fuel is required. And the plant can be tailored to produce enough plutonium to meet its need or to provide fuel for other nuclear plants

The safety basis of the integral fast reactor program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

Benchmark physics tests in the metallic-fuelled assembly ZPPR-15

International Nuclear Information System (INIS)

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1987-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size, liquid metal reactor are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics and spectrum. Analysis was done with 3-D nodal diffusion calculations and ENDFIB-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

Melcor benchmarking against integral severe fuel damage tests

Energy Technology Data Exchange (ETDEWEB)

Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

1995-09-01

MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

Science.gov (United States)

Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

2013-07-01

A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

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

Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

International Nuclear Information System (INIS)

Casal, V.; Graf, E.; Hartmann, W.

1976-09-01

The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

Results from the characterisation of the Futurix-FTA metal alloy transmutation fuels

International Nuclear Information System (INIS)

Rory Kennedy, J.; O'Holleran, Th.; Keiser, D.

2007-01-01

Full text of publication follows. Idaho National Laboratory has been developing and irradiation testing a number of fuels and fuel types for actinide transmutation as part of the Advanced Fuel Cycle Initiative (AFCI). Fuel types under consideration include both fertile (fast reactor systems) and fertile-free (accelerator-driven systems) metallic alloys. Most recently, fuel fabrication was completed and the fuel pins shipped to the fast flux Phenix reactor in Marcoule, France for irradiation testing as part of the FUTURIX-FTA experiment: an international experiment involving the USA, France, the European Commission and Japan. The metal alloy fuels for this experiment are the low-fertile U-29Pu-4Am-2Np-30Zr and the non-fertile Pu-12Am-40Zr. The fresh fuels have been fully characterised for chemical composition, phase, microstructure, thermal behaviour and fuel-cladding-chemical-interaction (FCCI). Preliminary FCCI results raised some safety concerns with respect to the formation of low melting phases and cladding degradation, which could preclude a fuel from consideration. Results from diffusion couple experiments between the non-fertile fuel Pu-12Am-40Zr and the ferritic HT9 and 422 stainless steels (SS) used in the AFC experiments in the ATR reactor (USA) compared to the austenitic AIM1 SS used in the FUTURIX-FTA experiments in the Phenix reactor (France) indicate significant inter-diffusion with the AIM1 SS. Up to about a 30-fold increase in the diffusion of iron (and accompanying Ni and Cr) into the fuel at 650 C was observed compared to the 422 SS studies. Comparable studies between the low-fertile U-29Pu-4Am-2Np-30Zr fuel alloy and the AIM1 SS show virtually no inter-diffusion. The Fe (along with small amounts of Ni and Cr) appears as small precipitates in the fuel alloy with only minor concentrations identified in the fuel alloy matrix. These results will be discussed in terms of mechanisms of the inter-diffusion and the difference in behaviour between the

Development of integrated cask body and base plate

International Nuclear Information System (INIS)

Sasaki, T.; Koyama, Y.; Yoshida, T.; Wada, T.

2015-01-01

The average of occupancy of stored spent-fuel in the nuclear power plants have reached 70 percent and it is anticipated that the demand of metal casks for the storage and transportation of spent-fuel rise after resuming the operations. The main part of metal cask consists of main body, neutron shield and external cylinder. We have developed the manufacturing technology of Integrated Cask Body and Base Plate by integrating Cask Body and Base Plate as monolithic forging with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Here, we report the manufacturing technology, code compliance and obtained properties of Integrated Cask body and Base Plate. (author)

Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

Energy Technology Data Exchange (ETDEWEB)

Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

2016-11-15

The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

International Nuclear Information System (INIS)

Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

1978-01-01

Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

Science.gov (United States)

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

2009-12-22

The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

Assessment of heavy metal contamination of dust at some selected fuel filling stations in Accra

International Nuclear Information System (INIS)

Afrifa, C. G.

2011-07-01

Heavy metal contaminated dust particles of fuel filling stations can be re-suspended into the ambient air and serve as a source of atmospheric pollution since the fine particles are aerodynamic and have longer life time in ambient air. This can cause ill-health effect on the fuel attendants and residents within the neighbourhood especially infants and the aged who are more vulnerable. In spite of this, not much research has been done on heavy metal contamination of dust at fuel filling stations. In this study, 55 dust samples were collected from six fuel filling stations in the Ga-East district and Accra Metropolitan assembly, both in Accra, in order to assess the levels of contamination of heavy metals; their possible sources and the human health risk associated with them. The dust samples were divided into two parts with one part sieved into four fractions using metric mesh sizes 500 µm, 200µm, 100µm and 45 µm, and pulverised. Total concentrations of heavy metals (Si, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Zn, Ga, Br, Rb, Sr, Y, Zr and Pb) were determined in the dust samples using energy dispersive X-ray fluorescence analysis. The pollution indices; enrichment factors (EF), index of geoaccumulation (Igeo), contamination factor (CF) and pollution load index (PLI) were used to identify possible levels of pollution from anthropogenic sources. The possible sources of metals were also identified with principal component analysis. Noncancer effect of children and adults due to exposure to dust from these fuel filling stations were also estimated. For the three fuelling areas, the average concentrations of V, Cr, Ni and Cu exceeded the acceptable values in common soil in the <45 µm fraction. The average concentration of Zn however exceeded the acceptable value only at the mixed-fuel fuelling area whereas the average concentration of Pb was within the acceptable value for all three fuelling areas. The dust samples showed moderate to significant enrichments for V, Cu, Br

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

Advances by the Integral Fast Reactor Program

International Nuclear Information System (INIS)

Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

1991-01-01

The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

Fuel failure monitoring system design approach for KALIMER

International Nuclear Information System (INIS)

Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

1998-01-01

Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

Integral benchmarks with reference to thorium fuel cycle

International Nuclear Information System (INIS)

Ganesan, S.

2003-01-01

This is a power point presentation about the Indian participation in the CRP 'Evaluated Data for the Thorium-Uranium fuel cycle'. The plans and scope of the Indian participation are to provide selected integral experimental benchmarks for nuclear data validation, including Indian Thorium burn up benchmarks, post-irradiation examination studies, comparison of basic evaluated data files and analysis of selected benchmarks for Th-U fuel cycle

Catalyst development and systems analysis of methanol partial oxidation for the fuel processor - fuel cell integration

Energy Technology Data Exchange (ETDEWEB)

Newson, E; Mizsey, P; Hottinger, P; Truong, T B; Roth, F von; Schucan, Th H [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1999-08-01

Methanol partial oxidation (pox) to produce hydrogen for mobile fuel cell applications has proved initially more successful than hydrocarbon pox. Recent results of catalyst screening and kinetic studies with methanol show that hydrogen production rates have reached 7000 litres/hour/(litre reactor volume) for the dry pox route and 12,000 litres/hour/(litre reactor volume) for wet pox. These rates are equivalent to 21 and 35 kW{sub th}/(litre reactor volume) respectively. The reaction engineering problems remain to be solved for dry pox due to the significant exotherm of the reaction (hot spots of 100-200{sup o}C), but wet pox is essentially isothermal in operation. Analyses of the integrated fuel processor - fuel cell systems show that two routes are available to satisfy the sensitivity of the fuel cell catalysts to carbon monoxide, i.e. a preferential oxidation reactor or a membrane separator. Targets for individual system components are evaluated for the base and best case systems for both routes to reach the combined 40% efficiency required for the integrated fuel processor - fuel cell system. (author) 2 figs., 1 tab., 3 refs.

Criticality safety evaluation of the fuel cycle facility electrorefiner

International Nuclear Information System (INIS)

Lell, R.M.; Mariani, R.D.; Fujita, E.K.; Benedict, R.W.; Turski, R.B.

1993-01-01

The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis

Enhanced Endosomal Escape by Light-Fueled Liquid-Metal Transformer.

Science.gov (United States)

Lu, Yue; Lin, Yiliang; Chen, Zhaowei; Hu, Quanyin; Liu, Yang; Yu, Shuangjiang; Gao, Wei; Dickey, Michael D; Gu, Zhen

2017-04-12

Effective endosomal escape remains as the "holy grail" for endocytosis-based intracellular drug delivery. To date, most of the endosomal escape strategies rely on small molecules, cationic polymers, or pore-forming proteins, which are often limited by the systemic toxicity and lack of specificity. We describe here a light-fueled liquid-metal transformer for effective endosomal escape-facilitated cargo delivery via a chemical-mechanical process. The nanoscale transformer can be prepared by a simple approach of sonicating a low-toxicity liquid-metal. When coated with graphene quantum dots (GQDs), the resulting nanospheres demonstrate the ability to absorb and convert photoenergy to drive the simultaneous phase separation and morphological transformation of the inner liquid-metal core. The morphological transformation from nanospheres to hollow nanorods with a remarkable change of aspect ratio can physically disrupt the endosomal membrane to promote endosomal escape of payloads. This metal-based nanotransformer equipped with GQDs provides a new strategy for facilitating effective endosomal escape to achieve spatiotemporally controlled drug delivery with enhanced efficacy.

Viscosity Meaurement Technique for Metal Fuels

Energy Technology Data Exchange (ETDEWEB)

Ban, Heng [Utah State Univ., Logan, UT (United States). Mechanical and Aerospace Engineering; Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-02-09

Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, the most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.

Viscosity Meaurement Technique for Metal Fuels

International Nuclear Information System (INIS)

Ban, Heng

2015-01-01

Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, the most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Time constants and transfer functions for a homogeneous 900 MWt metallic fueled LMR

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1988-01-01

Nodal transfer functions are calculated for a 900 MWt U10Zr-fueled sodium cooled reactor. From the transfer functions the time constants, feedback reactivity transfer function coefficients, and power coefficients can be determined. These quantities are calculated for core fuel, upper and lower axial reflector steel, radial blanket fuel, radial reflector steel, and B 4 C rod shaft expansion effect. The quantities are compared to the analogous quantities of a 60 MWt metallic-fueled sodium cooled Experimental Breeder Reactor II configuration. 8 refs., 2 figs., 6 tabs

Flow channel shape optimum design for hydroformed metal bipolar plate in PEM fuel cell

Energy Technology Data Exchange (ETDEWEB)

Peng, Linfa; Lai, Xinmin; Liu, Dong' an; Hu, Peng [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China); Ni, Jun [Department of Mechanical Engineering and Applied Mechanics, University of Michigan, Ann Arbor, MI 48109 (United States)

2008-03-15

Bipolar plate is one of the most important and costliest components of polymer electrolyte membrane (PEM) fuel cells. Micro-hydroforming is a promising process to reduce the manufacturing cost of PEM fuel cell bipolar plates made of metal sheets. As for hydroformed bipolar plates, the main defect is the rupture because of the thinning of metal sheet during the forming process. The flow channel section decides whether high quality hydroformed bipolar plates can be successively achieved or not. Meanwhile, it is also the key factor that is related with the reaction efficiency of the fuel cell stacks. In order to obtain the optimum flow channel section design prior the experimental campaign, some key geometric dimensions (channel depth, channel width, rib width and transition radius) of flow channel section, which are related with both reaction efficiency and formability, are extracted and parameterized as the design variables. By design of experiments (DOE) methods and an adoptive simulated annealing (ASA) optimization method, an optimization model of flow channel section design for hydroformed metal bipolar plate is proposed. Optimization results show that the optimum dimension values for channel depth, channel width, rib width and transition radius are 0.5, 1.0, 1. 6 and 0.5 mm, respectively with the highest reaction efficiency (79%) and the acceptable formability (1.0). Consequently, their use would lead to improved fuel cell efficiency for low cost hydroformed metal bipolar plates. (author)

Effects of accelerated degradation on metal supported thin film-based solid oxide fuel cell

DEFF Research Database (Denmark)

Reolon, R. P.; Sanna, S.; Xu, Yu

2018-01-01

A thin film-based solid oxide fuel cell is deposited on a Ni-based metal porous support by pulsed laser deposition with a multi-scale-graded microstructure design. The fuel cell, around 1 μm in thickness, is composed of a stabilized-zirconia/doped-ceria bi-layered dense electrolyte and nanostruct......A thin film-based solid oxide fuel cell is deposited on a Ni-based metal porous support by pulsed laser deposition with a multi-scale-graded microstructure design. The fuel cell, around 1 μm in thickness, is composed of a stabilized-zirconia/doped-ceria bi-layered dense electrolyte......, electrochemical performances are steady, indicating the stability of the cell. Under electrical load, a progressive degradation is activated. Post-test analysis reveals both mechanical and chemical degradation of the cell. Cracks and delamination of the thin films promote a significant nickel diffusion and new...

Development of metal uranium fuel and testing of construction materials (I-VI); Part I

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

The data-base of properties of actinides for metal fuels

International Nuclear Information System (INIS)

Inoue, Tadashi; Kurata, Masateru

1989-01-01

It is developed the technology that transuranium elements (TRUs) to be recovered from high active wastes transmute into relatively short lived nuclides by burning them within metallic fuel alloys. In this paper, we collect published data of properties of TRUs and U-Pu(-Zr) alloys and make up the data base for the design study of alloys with TRUs. In addition, the data base possesses a function of statistic analysis in order to facilitate the comparison of data and can afford to estimate properties. This data base collects (1) properties affecting fuel temperature and microstructure, (2) mechanical properties and (3) fundamental properties such as hardness and density, and furthermore, (1) fission gas release, (2) swelling and (3) fuel-cladding interaction and eutectic property as irradiation behavior. (author)

Industrial integration of the fuel cycle in Argentina

International Nuclear Information System (INIS)

Koll, J.H.; Kittl, J.E.; Parera, C.A.; Coppa, R.C.; Aguirre, E.J.

1977-01-01

The paper describes the power reactor construction programme in Argentina for the period 1976-1985, on the basis of which the nuclear fuel requirements have been determined. Activities connected with the fuel cycle commenced in 1950 in Argentina with the prospection and working of uranium deposits. On the basis of the nuclear power programme described, plans have been drawn up for the establishment of the industrial plants that will be needed to ensure the domestic supply of fuel. The demand for fuel is correlated with the availability of uranium resources and it is shown to be desirable from the technical, economic and industrial point of view to integrate the front end of the fuel cycle, keeping the irradiation aspects and the tail end at the development level. The authors report the progress that has been made in this field and describe current programmes covering prospection, concentrate production, nuclear purification, conversion to uranium dioxide, production of special alloys and fuel element fabrication. (author)

Industrial integration of the fuel cycle in Argentina

International Nuclear Information System (INIS)

Koll, J.H.; Kittl, J.E.; Parera, C.A.; Coppa, R.C.; Aguirre, E.J.

1983-01-01

The power-reactor construction program in Argentina for the period 1976-1985 is described on the basis of which the nuclear-fuel requirements have been determined. Activities connected with the fuel cycle commenced in 1950 in Argentina with the prospection and working of uranium deposits. On the basis of the nuclear power program described, plans have been drawn up for the establishment of the industrial plants that will be needed to ensure the domestic supply of fuel. The demand for fuel is correlated with the availability of uranium resoures and it is shown to be desirable from the technical, economic and industrial point of view to integrate the front end of the fuel cycle, keeping the irradiation aspects and the tail end at the development level. Progress made in this field and current programs covering exploration, concentrate production, nuclear purification, conversion to uranium dioxide, production of special alloys and fuel element fabrication are described

Role of metal/silicon semiconductor contact engineering for enhanced output current in micro-sized microbial fuel cells

KAUST Repository

Mink, Justine E.

2013-11-25

We show that contact engineering plays an important role to extract the maximum performance from energy harvesters like microbial fuel cells (MFCs). We experimented with Schottky and Ohmic methods of fabricating contact areas on silicon in an MFC contact material study. We utilized the industry standard contact material, aluminum, as well as a metal, whose silicide has recently been recognized for its improved performance in smallest scale integration requirements, cobalt. Our study shows that improvements in contact engineering are not only important for device engineering but also for microsystems. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

Energy Technology Data Exchange (ETDEWEB)

Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

1979-09-30

The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

Improved nuclear fuel element

International Nuclear Information System (INIS)

Klepfer, H.H.

1974-01-01

A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

International Nuclear Information System (INIS)

Hu Yun; Xu Mi; Wang Kan

2010-01-01

In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

Design and integration of a hydrogen storage on metallic hydrides

International Nuclear Information System (INIS)

Botzung, M.

2008-01-01

This work presents a hydrogen storage system using metal hydrides for a Combined Heat and Power (CHP) system. Hydride storage technology has been chosen due to project specifications: high volumetric capacity, low pressures (≤ 3.5 bar) and low temperatures (≤ 75 C: fuel cell temperature). During absorption, heat from hydride generation is dissipated by fluid circulation. An integrated plate-fin type heat exchanger has been designed to obtain good compactness and to reach high absorption/desorption rates. At first, the storage system has been tested in accordance with project specifications (absorption 3.5 bar, desorption 1.5 bar). Then, the hydrogen charge/discharge times have been decreased to reach system limits. System design has been used to simulate thermal and mass comportment of the storage tank. The model is based on the software Fluent. We take in consideration heat and mass transfers in the porous media during absorption/desorption. The hydride thermal and mass behaviour has been integrated in the software. The heat and mass transfers experimentally obtained have been compared to results calculated by the model. The influence of experimental and numerical parameters on the model behaviour has also been explored. (author) [fr

The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

Energy Technology Data Exchange (ETDEWEB)

Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

International Nuclear Information System (INIS)

Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

2016-01-01

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

Fuel pin integrity assessment under large scale transients

International Nuclear Information System (INIS)

Dutta, B.K.

2006-01-01

The integrity of fuel rods under normal, abnormal and accident conditions is an important consideration during fuel design of advanced nuclear reactors. The fuel matrix and the sheath form the first barrier to prevent the release of radioactive materials into the primary coolant. An understanding of the fuel and clad behaviour under different reactor conditions, particularly under the beyond-design-basis accident scenario leading to large scale transients, is always desirable to assess the inherent safety margins in fuel pin design and to plan for the mitigation the consequences of accidents, if any. The severe accident conditions are typically characterized by the energy deposition rates far exceeding the heat removal capability of the reactor coolant system. This may lead to the clad failure due to fission gas pressure at high temperature, large- scale pellet-clad interaction and clad melting. The fuel rod performance is affected by many interdependent complex phenomena involving extremely complex material behaviour. The versatile experimental database available in this area has led to the development of powerful analytical tools to characterize fuel under extreme scenarios

Trace metal assay of uranium silicide fuel

International Nuclear Information System (INIS)

Kulkarni, M.J.; Argekar, A.A.; Thulasidas, S.K.; Dhawale, B.A.; Rajeswari, B.; Adya, V.C.; Purohit, P.J.; Neelam, G.; Bangia, T.R.; Page, A.G.; Sastry, M.D.; Iyer, R.H.

1994-01-01

A comprehensive trace metal assay of uranium silicide, a fuel for nuclear research reactors that employs low-enrichment uranium, is carried out by atomic spectrometry. Of the list of specification elements, 21 metallic elements are determined by a direct current (dc) arc carrier distillation technique; the rare earths yttrium and zirconium are chemically separated from the major matrix followed by a dc arc/inductively coupled argon plasma (ICP) excitation technique in atomic emission spectrometry (AES); silver is determined by electrothermal atomization-atomic absorption spectrometry (ETA-AAS) without prior chemical separation of the major matrix. Gamma radioactive tracers are used to check the recovery of rare earths during the chemical separation procedure. The detection limits for trace metallics vary in the 0.1- to 40-ppm range. The precision of the determinations as evaluated from the analysis of the synthetic sample with intermediate range analyte concentration is better than 25% relative standard deviation (RSD) for most of the elements employing dc arc-AES, while that for silver determination by ETS-AAS is 10% RSD. The precision of the determinations for four crucially important rare earths by ICP-AES is better than 3% RSD

Function of all-metal separators for waste fuels. Phase 1; Funktion av allmetallseparatorer foer avfallsbraenslen. Etapp 1

Energy Technology Data Exchange (ETDEWEB)

Jacoby, Juergen; Wrangensten, Lars

2004-08-01

Various waste incineration facilities, which use different types of waste fuels, have difficulties with a high content of non-magnetic metal, especially aluminum in their fuels. Aluminum may melt on the grate and can lead to corrosion or fouling in the furnace. Additionally, a high content of aluminum in the flyash may cause difficulties in terms of storage or further use of the ash as e.g. construction material. The industrial demand for efficient separators for non-magnetic metals from a fuel stream is rather large. There is however some uncertainty in the performance and efficiency of metal separators. Two types of separators can be found, the first type is called eddy current separator, the other type is based upon a metal detector with a sorting unit in the form of a chute or similar afterwards. An eddy current separator consists of a fast rotating drum containing several permanent magnets with alternating polarity. Due to the rotation, the change in the magnetic field induces eddy currents in conducting materials. The eddy currents cause a force in non-magnetic metal, the Lorentz force, which repels the material away from the rotating drum while all other material follows the systems flow direction. Systems equipped with a metal detector activate a mechanical sorting device, separate chute or air nozzles, when a metal particle is detected. In contrast to eddy current separators all types of metals can be detected and sorted out by systems based on metal detector. Several technical solutions for metal separation supplied by various manufacturers are described in the report. The companies have been asked to supply product information on the working principle, technical data, efficiency and limits for different types of metals. Two reference power plants have been visited and their experiences with all-metal separators are described. Haendeloeverket in Norrkoeping uses eddy current separators for separation of non-magnetic metals from household waste

Nuclear criticality safety program at the Fuel Cycle Facility

International Nuclear Information System (INIS)

Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

1994-01-01

The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

Carbon-Nanotube-Supported Bio-Inspired Nickel Catalyst and Its Integration in Hybrid Hydrogen/Air Fuel Cells.

Science.gov (United States)

Gentil, Solène; Lalaoui, Noémie; Dutta, Arnab; Nedellec, Yannig; Cosnier, Serge; Shaw, Wendy J; Artero, Vincent; Le Goff, Alan

2017-02-06

A biomimetic nickel bis-diphosphine complex incorporating the amino acid arginine in the outer coordination sphere was immobilized on modified carbon nanotubes (CNTs) through electrostatic interactions. The functionalized redox nanomaterial exhibits reversible electrocatalytic activity for the H 2 /2 H + interconversion from pH 0 to 9, with catalytic preference for H 2 oxidation at all pH values. The high activity of the complex over a wide pH range allows us to integrate this bio-inspired nanomaterial either in an enzymatic fuel cell together with a multicopper oxidase at the cathode, or in a proton exchange membrane fuel cell (PEMFC) using Pt/C at the cathode. The Ni-based PEMFC reaches 14 mW cm -2 , only six-times-less as compared to full-Pt conventional PEMFC. The Pt-free enzyme-based fuel cell delivers ≈2 mW cm -2 , a new efficiency record for a hydrogen biofuel cell with base metal catalysts. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Development of comprehensive long-term-dry stored Spent Fuel INtegrity EvaLuator [SFINEL] - I

International Nuclear Information System (INIS)

Kwon, H. M.; Yang, Y. S.; Kim, Y. S.; You, K. S.; Min, D. K.; No, S. K.

1999-01-01

Safe management of spent nuclear fuels is socially, technically, and economically very important in terms of environmental protection and utilization of recyclable resources. One of the most critical parts in the management is to establish the comprehensive monitoring system which can maintain and confirm the integrity of the spent fuels, whenever necessary, until final policy is determined on the their treatment and disposal. Especially in the first stage of maturing up the system, it is essential to secure a computing tool or code which can evaluate the integrity of the fuel cladding based on its power history and cladding degradation mechanisms. SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed in this research. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms

Material accountancy for metallic fuel pin casting

International Nuclear Information System (INIS)

Bucher, R.G.; Orechwa, Y.; Beitel, J.C.

1995-01-01

The operation of the Fuel Conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. The pin casting operation, although only one of several operations in FCF, was the first to be on-line. As such, it has served to demonstrate the material accountancy system in many of its facets. This paper details, for the operation of the pin casting process with depleted uranium, the interaction between the mass tracking system (MTG) and some of the ancillary computer codes which generate pertinent information for operations and material accountancy. It is necessary to distinguish between two types of material balance calculations -- closeout for operations and material accountancy for safeguards. The two have much in common, for example, the mass tracking system database and the calculation of an inventory difference, but, in general, are not congruent with regard to balance period and balance spatial domain. Moreover, the objective, assessment, and reporting requirements of the calculated inventory difference are very different in the two cases

Performance of Zr as FCCI barrier layer for metallic fuel of fast reactor

International Nuclear Information System (INIS)

Kaity, Santu; Bhagat, R.K.; Kutty, T.R.G.; Kumar, Arun; Laik, A.; Kamath, H.S.

2011-01-01

Uranium-plutonium (U-Pu) and uranium-plutonium-zirconium (U-Pu-Zr) alloys have been considered as promising advanced fuels for fast reactor in India because of its high breeding potential, high thermal conductivity, high fissile and fertile atom densities, low doubling time and ease of fabrication compared to other ceramic fuels. The chemical compatibility between the fuel and clad material also known as fuel-clad chemical interaction (FCCI) has been recognized as one of the major concerns about the performance of the metallic fuel. Primarily, two design concepts have been proposed for the metallic fuel development programme for FBRs. One of them is based on sodium bonded ternary U-Pu-Zr alloy with T91 grade steel clad, and the other consists of binary U-Pu alloy mechanically bonded to T91 clad with a Zr liner between the fuel and clad. U will be the axial blanket material for U-Pu binary fuel. In the present investigation, the performance of Zr as FCCI barrier layer was studied through diffusion couple experiments of U/Zr/T91. A thin Zr foil (thickness ∼ 200 μm) sandwiched between U and T91 discs was kept inside a fixture made of Inconel 600 alloy. The fixture was encapsulated in quartz tube under Helium atmosphere and then heated at 650, 700 and 750 deg C for upto 1500 h. The extent of reaction and composition of phases formed were analyzed by scanning electron microscope (SEM) equipped with an energy dispersive spectrometer (EDS) and electron probe microanalyser (EPMA) equipped with wavelength dispersive spectrometer (WDS)

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Integral Fast Reactor Program

International Nuclear Information System (INIS)

Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

1993-06-01

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

Species redistribution during solidification of nuclear fuel waste metal castings

Energy Technology Data Exchange (ETDEWEB)

Naterer, G F; Schneider, G E [Waterloo Univ., ON (Canada)

1994-12-31

An enthalpy-based finite element model and a binary system species redistribution model are developed and applied to problems associated with solidification of nuclear fuel waste metal castings. Minimal casting defects such as inhomogeneous solute segregation and cracks are required to prevent container corrosion and radionuclide release. The control-volume-based model accounts for equilibrium solidification for low cooling rates and negligible solid state diffusion for high cooling rates as well as intermediate conditions. Test problems involving nuclear fuel waste castings are investigated and correct limiting cases of species redistribution are observed. (author). 11 refs., 1 tab., 13 figs.

Metal Phosphates as Proton Conducting Materials for Intermediate Temperature Fuel Cell and Electrolyser Applications

DEFF Research Database (Denmark)

Anfimova, Tatiana

The present thesis presents the results achieved during my ph.d. project on a subject of intermediate temperature proton conducting metal phosphates as electrolyte materials for fuel cells and electrolysers. Fuel cells and electrolysers are electrochemical devices with high energy conversion...... with a proton conductivity of above 10-2S cm-1. Chapter 1 of the thesis is an introduction to basics of fuel cell and electrolyser technologies as well as proton conducting materials. Extended discussion on the proton conducting materials, a particularly phosphates is made in Chapter 2. Three major types...... starts with synthesis and investigation of three rare earth metal phosphate hydrates, which is first presented in Chapter 5. Structural and surface water as well as its stability has been investigated using thermogravimetric and differential thermal analyses combined with structural modeling calculations...

The Integral Fast Reactor

International Nuclear Information System (INIS)

Chang, Y.I.