Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

International Nuclear Information System (INIS)

Prince, B.E.; Hadley, S.W.

1983-01-01

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of the time-dependent aspects of plutonium requirements and supply relative to this transition

The dupic fuel cycle synergism between LWR and HWR

International Nuclear Information System (INIS)

Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.

1999-01-01

The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)

Flexible fuel cycle system for the transition from LWR to FBR

International Nuclear Information System (INIS)

Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Sasahira, Akira; Inoue, Tadashi; Minato, Kazuo; Sato, Seichi

2009-01-01

Japan will deploy commercial fast breeder reactor (FBR) from around 2050 under the suitable conditions for the replacement of light water reactor (LWR) with FBR. The transition scenario from LWR to FBR is investigated in detail and the flexible fuel cycle initiative (FFCI) system has been proposed as a optimum transition system. The FFCI removes ∼95% uranium from LWR spent fuel (SF) in LWR reprocessing and residual material named Recycle Material (RM), which is ∼1/10 volume of original SF and contains ∼50% U, ∼10% Pu and ∼40% other nuclides, is treated in FBR reprocessing to recover Pu and U. If the FBR deployment speed becomes lower, the RM will be stored until the higher speed again. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR SF during the transition period. The economy is better for FFCI due to the smaller LWR reprocessing facility (no Pu/U recovery and fabrication). The FFCI can supply high Pu concentration RM, which has high proliferation resistance and flexibly respond to FBR introduction rate changes. Volume minimization of LWR SF is possible for FFCI by its conversion to RM. Several features of FFCI were quantitatively evaluated such as Pu mass balance, reprocessing capacities, LWR SF amounts, RM amounts, and proliferation resistance to compare the effectiveness of the FFCI system with other systems. The calculated Pu balance revealed that the FFCI could supply enough but no excess Pu to FBR. These evaluations demonstrated the applicability of FFCI system to the transition period from LWR to FBR cycles. (author)

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-15

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U₃Si₂ as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U₃Si₂ fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U₃Si₂. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

Effects of cooling time on a closed LWR fuel cycle

International Nuclear Information System (INIS)

Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

2012-01-01

In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

International Nuclear Information System (INIS)

Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

2008-01-01

Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

Modular approach to LWR in-core fuel management

International Nuclear Information System (INIS)

Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

1980-01-01

The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

Fission product release from high gap-inventory LWR fuel under LOCA conditions

International Nuclear Information System (INIS)

Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

1980-01-01

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

Development of top nozzle for Korean standard LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, S. K.; Kim, I. K.; Choi, K. S.; Kim, Y. H.; Lee, J. N.; Kim, H. K. [KNFC, Taejon (Korea, Republic of)

2001-10-01

Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles.

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Hiramoto, K.; Maru, A.

1984-01-01

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

2009-07-01

The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2000-11-01

A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-08-29

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U₃Si₂ at LWR conditions. The fission gas behavior of U₃Si₂ can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U₃Si₂ for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U₃Si₂ at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U₃Si₂ as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

Economic analyses of LWR fuel cycles

International Nuclear Information System (INIS)

Field, F.R.

1977-05-01

An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled

Tools for LWR spent fuel characterization: Assembly classes and fuel designs

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1991-01-01

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

Equipment designs for the spent LWR fuel dry storage demonstration

International Nuclear Information System (INIS)

Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

1980-01-01

In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

Validating the BISON fuel performance code to integral LWR experiments

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

2016-05-15

Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

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

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

International Nuclear Information System (INIS)

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

1992-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-10-01

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

Nondestructive evaluation of LWR spent fuel shipping casks

International Nuclear Information System (INIS)

Ballard, D.W.

1978-02-01

An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined

International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

International Nuclear Information System (INIS)

Sowder, Andrew

2013-01-01

Following the March 2011 multi-unit accident at the Fukushima Daiichi plant, there has been increased interest in the development of breakthrough nuclear fuel designs that can reduce or eliminate many of the outcomes of a severe accident at a light water reactor (LWR) due to loss of core cooling following an extended station blackout or other initiating event. With this interest and attention comes a unique opportunity for the nuclear industry to fundamentally change the nature and impact of severe accidents. Clearly, this is no small feat. The challenges are many and the technical barriers are high. Early estimates for moving maturing R and D concepts to the threshold of commercialisation exceed one billion USD. Given the anticipated effort and resources required, no single entity or group can succeed alone. Accordingly, the Electric Power Research Institute (EPRI) sees the need for and promise of cooperation among many stakeholders on an international scale to bring about what could be transformation in LWR fuel performance and robustness. An important initial task in any R and D programme is to define the goals and metrics for measuring success. As starting points for accident-tolerant fuel development, the extension of core coolability under loss of coolant conditions and the elimination or reduction of hydrogen generation are widely recognised R and D endpoints for deployment. Furthermore, any new LWR fuel technology will, at a minimum, need to (1) be compatible with the safe, economic operation of existing plants and (2) maintain acceptable or improve nuclear fuel performance under normal operating conditions. While the primary focus of R and D to date has been on cladding and fuel improvements, there are a number of other potential paths to improve outcomes following a severe accident at an LWR that include modifications to other fuel hardware and core internals to fully address core coolability, criticality, and hydrogen generation concerns. The US

NUPEC proves reliability of LWR fuel assemblies

International Nuclear Information System (INIS)

Anon.

1987-01-01

It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

International Nuclear Information System (INIS)

Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

1986-09-01

The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

Safety criteria related to microheterogeneities in LWR mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Mostin, N.

1978-01-01

The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

Pie technique of LWR fuel cladding fracture toughness test

International Nuclear Information System (INIS)

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

2006-01-01

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

Standard casks for the transport of LWR spent fuel. Storage/transport casks for long cooled spent fuel

International Nuclear Information System (INIS)

Blum, P.; Sert, G.; Gagnon, R.

1983-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks are presently used for European and intercontinental transports and manufactured under TRANSNUCLEAIRE supervision in different countries. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardization faciliting fabrication, operation and spare part supply. Recently, TRANSNUCLEAIRE also developed a new generation of casks for the dry storage and occasional transport of LWR spent fuel which has been cooled for 5 years or 7 years in case of consolidated fuel rods. These casks have an optimum payload which takes into account the shielding requirements and the weight limitations at most sites. This paper deals more particularly with the TN 24 model which exists in 4 versions among which one for 24 PWR 900 fuel assemblies and another one for the consolidated fuel rods from 48 of same fuel assemblies

Criticality impacts on LWR fuel storage efficiency

International Nuclear Information System (INIS)

Napolitano, D.

1992-01-01

This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Analysis of alternative light water reactor (LWR) fuel cycles

International Nuclear Information System (INIS)

Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.

1979-12-01

Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails

Conceptual design of a spent LWR fuel recycle complex

International Nuclear Information System (INIS)

Kirk, B.H.

1980-01-01

Purpose was to design a licensable facility, to make cost-benefit analyses of alternatives, and to aid in developing licensing criteria. The Savannah River Plant was taken to be the site for the recycle complex. The spent LWR fuel will be processed through the plant at the rate of 3000 metric tons of heavy metal per year. The following aspects of the complex are discussed: operation, maintenance, co-conversion (Coprecal), waste disposal, off-gas treatment, ventilation, safeguards, accounting, equipment and fuel fabrication. Differences between the co-processing case and the separated streams case are discussed. 44 figures

Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

International Nuclear Information System (INIS)

Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

1983-09-01

Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

Data needs for long-term dry storage of LWR fuel. Interim report

International Nuclear Information System (INIS)

Einziger, R.E.; Baldwin, D.L.; Pitman, S.G.

1998-04-01

The NRC approved dry storage of spent fuel in an inert environment for a period of 20 years pursuant to 10CFR72. However, at-reactor dry storage of spent LWR fuel may need to be implemented for periods of time significantly longer than the NRC's original 20-year license period, largely due to uncertainty as to the date the US DOE will begin accepting commercial spent fuel. This factor is leading utilities to plan not only for life-of-plant spent-fuel storage during reactor operation but also for the contingency of a lengthy post-shutdown storage. To meet NRC standards, dry storage must (1) maintain subcriticality, (2) prevent release of radioactive material above acceptable limits, (3) ensure that radiation rates and doses do not exceed acceptable limits, and (4) maintain retrievability of the stored radioactive material. In light of these requirements, this study evaluates the potential for storing spent LWR fuel for up to 100 years. It also identifies major uncertainties as well as the data required to eliminate them. Results show that the lower radiation fields and temperatures after 20 years of dry storage promote acceptable fuel behavior and the extension of storage for up to 100 years. Potential changes in the properties of dry storage system components, other than spent-fuel assemblies, must still be evaluated

Preliminary concepts for detecting national diversion of LWR spent fuel

International Nuclear Information System (INIS)

Sonnier, C.S.; Cravens, M.N.

1978-04-01

Preliminary concepts for detecting national diversion of LWR spent fuel during storage, handling and transportation are presented. Principal emphasis is placed on means to achieve timely detection by an international authority. This work was sponsored by the Department of Energy/Office of Safeguards and Security (DOE/OSS) as part of the overall Sandia Fixed Facility Physical Protection Program

The necessity of improvement for the current LWR fuel assembly homogenization method

International Nuclear Information System (INIS)

Tang Chuntao; Huang Hao; Zhang Shaohong

2007-01-01

When the modern LWR core analysis method is used to do core nuclear design and in-core fuel management calculation, how to accurately obtain the fuel assembly homogenized parameters is a crucial issue. In this paper, taking the NEA C5G7-MOX benchmark problem as a severe test problem, which involves low-enriched uranium assemblies interspersed with MOX assemblies, we have re-examined the applicability of the two major assumptions of the modern equivalence theory for fuel assembly homoge- nization, i.e. the isolated assembly spatial spectrum assumption and the condensed two- group representation assumption. Numerical results have demonstrated that for LWR cores with strong spectrum interaction, both of these two assumptions are no longer applicable and the improvement for the homogenization method is necessary, the current two-group representation should be improved by the multigroup representation and the current reflective assembly boundary condition should be improved by the 'real' assembly boundary condition. This is a research project supported by National Natural Science Foundation of China (10605016). (authors)

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

Energy Technology Data Exchange (ETDEWEB)

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01

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

Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

2009-01-01

As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

Assesment On The Possibility To Modify Fabrication Equipment For Fabrication Of HWR And LWR Fuel Elements

International Nuclear Information System (INIS)

Tri-Yulianto

1996-01-01

Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable

Nondestructive evaluation of LWR spent fuel shipping casks

International Nuclear Information System (INIS)

Ballard, D.W.

1978-02-01

An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

International Nuclear Information System (INIS)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems

Safety-related investigations on power distribution in MOX fuel elements in LWR cores

International Nuclear Information System (INIS)

Kramer, E.; Langenbuch, S.

1991-01-01

For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

Recycling U and Pu in LWR

International Nuclear Information System (INIS)

Zheng Hualing.

1986-01-01

This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

International Nuclear Information System (INIS)

1987-12-01

This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1986-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr

Behavior of LWR fuel elements under accident conditions

International Nuclear Information System (INIS)

Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

1977-01-01

In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

International Nuclear Information System (INIS)

Sample, C.R.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

Energy Technology Data Exchange (ETDEWEB)

Sample, C R [comp.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

The concept of fuel cycle integrated molten salt reactor for transmuting Pu+MA from spent LWR fuels

International Nuclear Information System (INIS)

Hirose, Y.; Takashima, Y.

2001-01-01

Japan should need a new fuel cycle, not to save spent fuels indefinitely as the reusable resources but to consume plutonium and miner actinides orderly without conventional reprocessing. The key component is a molten salt reactor fueled with the Pu+MA (PMA) separated from LWR spent fuels using fluoride volatility method. A double-tiered once-through reactor system can burn PMA down to 5% remnant ratio, and can make PMA virtually free from the HAW to be disposed geometrically. A key issue to be demonstrated is the first of all solubility behavior of trifluoride species in the molten fuel salt of 7 LiF-BeF 2 mixture. (author)

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-12-10

Uranium silicide (U₃Si₂) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U₃Si₂, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U₃Si₂ at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U₃Si₂ in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U₃Si₂ at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U₃Si₂ fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U₃Si₂ fuel as an accident

Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

International Nuclear Information System (INIS)

Katayama, Y.B.

1979-04-01

Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

The Width of High Burnup Structure in LWR UO2 Fuel

International Nuclear Information System (INIS)

Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

2007-01-01

The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

The scale analysis sequence for LWR fuel depletion

International Nuclear Information System (INIS)

Hermann, O.W.; Parks, C.V.

1991-01-01

The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

Baseline descriptions for LWR spent fuel storage, handling, and transportation

Energy Technology Data Exchange (ETDEWEB)

Moyer, J.W.; Sonnier, C.S.

1978-04-01

Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

Baseline descriptions for LWR spent fuel storage, handling, and transportation

International Nuclear Information System (INIS)

Moyer, J.W.; Sonnier, C.S.

1978-04-01

Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

Spent fuel handling and storage facility for an LWR fuel reprocessing plant

International Nuclear Information System (INIS)

Baker, W.H.; King, F.D.

1979-01-01

The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-10-01

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

Energy Technology Data Exchange (ETDEWEB)

Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

2004-07-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

International Nuclear Information System (INIS)

Purcell, P.C.; Dallongeville, M.

2004-01-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2013-01-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

International Nuclear Information System (INIS)

Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

1990-08-01

A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

Decay heat and gamma dose-rate prediction capability in spent LWR fuel

International Nuclear Information System (INIS)

Neely, G.J.; Schmittroth, F.

1982-08-01

The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1985-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardisation facilitating fabrication, operation and spare part supply [fr

Integrity of neutron-absorbing components of LWR fuel systems

International Nuclear Information System (INIS)

Bailey, W.J.; Berting, F.M.

1991-03-01

A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs

New development in nondestructive measurement and verification of irradiated LWR fuels

International Nuclear Information System (INIS)

Lee, D.M.; Phillips, J.R.; Halbig, J.K.; Hsue, S.T.; Lindquist, L.O.; Ortega, E.M.; Caine, J.C.; Swansen, J.; Kaieda, K.; Dermendjiev, E.

1979-01-01

Nondestructive techniques for characterizing irradiated LWR fuel assemblies are discussed. This includes detection systems that measure the axial activity profile, neutron yield and gamma yield. A multi-element profile monitor has been developed that offers a significant improvement in speed and complexity over existing mechanical scanning systems. New portable detectors and electronics, applicable to safeguard inspection, are presented and results of gamma-ray and neutron measurements at commercial reactor facilities are given

Measurement and characterization of fission products released from LWR fuel

International Nuclear Information System (INIS)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

1984-01-01

Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

International Nuclear Information System (INIS)

Stout, R.B.; Merckx, K.R.; Holm, J.S.

1981-01-01

This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels

Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

International Nuclear Information System (INIS)

Gilbert, E.R.; Knox, C.A.; White, G.D.

1985-09-01

Interim dry storage of spent light-water reactor (LWR) fuel is being developed as a licensed technology in the United States. Because it is anticipated that license agreements will specify dry storage atmospheres, the behavior of spent LWR fuel in a nitrogen atmosphere during dry storage was investigated. In particular, the thermodynamics of reaction of nitrogen compounds (expected to form in the cover gas during dry storage) and residual impurities (such as moisture and oxygen) with Zircaloy cladding and with spent fuel at sites of cladding breaches were examined. The kinetics of reaction were not considered it was assumed that the 20 to 40 years of interim dry storage would be sufficient for reactions to proceed to completion. The primary thermodynamics reactants were found to be NO 2 , N 2 O, H 2 O 2 , and O 2 . The evaluation revealed that the limited inventories of these reactants produced by the source terms in hermetically sealed dry storage systems would be too low to cause significant spent fuel degradation. Furthermore, the oxidation of spent fuel to degrading O/U ratios is unlikely because the oxidation potential in moist nitrogen limits O/U ratios to values less than UO/sub 2.006/ (the equilibrium stoichiometric form in equilibrium with moist nitrogen). Tests were performed with bare spent UO 2 fuel and nonirradiated UO 2 pellets (with no Zircaloy cladding) in a nitrogen atmosphere containing moisture concentrations greater than encountered under dry storage conditions. These tests were performed for at least 1100 h at temperatures as high as 380 0 C, where oxidation reactions proceed in a matter of minutes. No visible degradation was detected, and weight changes were negligible

Comment: collection of assay data on isotopic composition in LWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

A study on the behavior of defected LWR spent fuel

International Nuclear Information System (INIS)

You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook

1992-03-01

To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)

LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

Energy Technology Data Exchange (ETDEWEB)

Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

1983-06-01

LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

Information on the evolution of severe LWR fuel element damage obtained in the CORA program

International Nuclear Information System (INIS)

Schanz, G.; Hagen, S.; Hofmann, P.; Sepold, L.; Schumacher, G.

1992-01-01

In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

International Nuclear Information System (INIS)

Bae, Gonghoon; Hong, Sergi

2012-01-01

The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

International Nuclear Information System (INIS)

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

International Nuclear Information System (INIS)

Whitesides, G.H.; Stephens, M.E.

1984-01-01

During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

Measurement and characterization of fission products released from LWR fuel

International Nuclear Information System (INIS)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

1984-01-01

Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

Estimation of the core-wide fuel rod damage during a LWR LOCA

International Nuclear Information System (INIS)

Mattila, L.; Sairanen, R.; Stengaard, J.-O.

1975-01-01

The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

International Nuclear Information System (INIS)

Mihalczo, J.T.

1987-01-01

The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

'CANDLE' burnup regime after LWR regime

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Nagata, Akito

2008-01-01

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

LWR-core behaviour project

International Nuclear Information System (INIS)

Paratte, J.M.

1982-07-01

The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

2013-01-01

Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

International Nuclear Information System (INIS)

Bomboni, Eleonora

2008-01-01

The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

Effect of long-term storage of LWR spent fuel on Pu-thermal fuel cycle

International Nuclear Information System (INIS)

Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; Itahara, Kuniyuki; Suzuki, Katsuo; Hamada, Koji

1998-01-01

According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu. Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO 2 spent fuels will be required. Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc. As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan. (author)

Qualification of the neutronic evolution of LWR fuels in MELUSINE

International Nuclear Information System (INIS)

Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.

1984-09-01

MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr

Performance and reliability of LWR fuel

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.

1977-01-01

The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted

Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

International Nuclear Information System (INIS)

Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

1998-01-01

This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

Quality assurance in the course of fabrication of LWR fuel

International Nuclear Information System (INIS)

Dressler, G.; Perry, J.A.

1982-01-01

A high quality level of LWR fuel elements can only be assured by a system of Quality Assurance measures purposefully designed, balanced, and appropriately applied. This includes application of and the appropriate balance between both system and product oriented measures. A prerequisite to the establishment of these measures is a precise analysis of the various influences of the individual process steps on the quality characteristics of the starting materials, semi-finished and finished products. In addition, these characteristics require classification criteria relative to their significance. The described classification is used to establish sampling plans and to disposition non-conformances. The EXXON Nuclear Quality Assurance system which is based on these principles is described and illustrated with some examples. (orig.)

Feasibility study on the development of advanced LWR fuel technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

Feasibility study on the development of advanced LWR fuel technology

Energy Technology Data Exchange (ETDEWEB)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

Energy profit ratio on LWR by uranium recycles

International Nuclear Information System (INIS)

Amano, Osamu; Uno, Takeki; Matsushima, Jun

2009-01-01

Energy profit ratio is defined as the ratio of output energy/input system total energy. In case of electric power generation, input energy is a total for fuel such as uranium mining and enrichment, fuel transportation, build nuclear power plant, M and O and for disposal waste and decommission of reactor vessel. Output energy is the total electricity on LWR during the plant life. EPR on both PWR and BWR is high value using gas centrifuge enrichment compared other type of electric power generation such as a thermal power, a hydraulic power, a wind power and a photovoltaic power. How is the EPR on LWR by MOX? We need understanding the energy of reprocessing spent fuel, MOX fuel fabrication, low level waste disposal and high level radioactive glass disposal. As we show the material balance for two cases, the first is the case of long term storage and reprocessing before FBR, the second is the MOX fuel cycle on LWR plant. The MOX fuel recycle is better EPR value rather than the case of long term storage and reprocessing before FBR (LTSRBF). At the gaseous diffusion enrichment case, MOX fuel recycle has 15 to 18% higher EPR value than LTSRBF. At the gas centrifuge enrichment case the MOX fuel recycle has 17 to 18 higher EPR value than LTSRBF. MOX fuel recycle decreases the uranium mining and refine mass, enrichment separative work and the spent fuel interim storage. It tells us the MOX fuel recycle is good way from view of EPR. (author)

A review on future trends of LWR fuel cycle costs

International Nuclear Information System (INIS)

Tamiya, S.; Otomo, T.; Meguro, T.

1977-01-01

In the cost estimations in the past, the main components of fuel cycle were mining and milling, uranium enrichment and fuel fabrication, and reprocessing charge deemed to be recovered by plutonium credit. Since the oil crisis, every component of fuel cycle cost has gone up in recent years as well as the construction cost of a power station. Recent analysis shows that the costs in the back end of fuel cycle are much higher than those anticipated several years ago, although their contribution to the electricity generating cost by nuclear would be small. The situation of the back end of the fuel cycle has been quite changed in recent years, and there are still many uncertainties in this field, that is, regulatory requirements for reprocessing plant such as safety, safeguards, environmental protection and high level waste management. So, it makes it more difficult to estimate the investment in this sector of fuel cycle, therefore, to estimate the cost of this sector. The institutional problems must be cleared in relation to the ultimate disposal of high level waste, too. Co-location of some parts of fuel cycle facilities may also affect on the fuel cycle costs. In this paper a review is made of the future trend of nuclear fuel cycle cost of LWR based on the recent analysis. Those factors which affect the fuel cycle costs are also discussed. In order to reduce the uncertainties of the cost estimations as soon as possible, the necessity is emphasized to discuss internationally such items as the treatment and disposal of high level radioactive wastes, siting issues of a reprocessing plant, physical protection of plutonium and the effects of plutonium on the environment

Long-term fuel cycle scenarios for advanced utilization of plutonium from LWRs

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji

2005-01-01

The Innovative Water Reactor for Flexible fuel cycle (FLWR) realizes multiple recycling and breeding of Pu, which enables effective utilization of the uranium resource, and is based on well-developed LWR technologies. This reactor offers flexibility for the future nuclear fuel cycle situation. Three scenarios were defined for future deployment of nuclear power generation and fuel cycle systems in Japan and analyzed from the view point of Pu recycle, natural uranium consumption and stock of spent fuels. The LWR with long-term Pu recycle with or without MOX fuel reprocessing needs uranium of about 9 thousands tons per year and accumulated uranium consumption of 1.5 million tons in 2150. If the FLWR with net conversion ratio of 0.89 and 1.04 would be introduced in 2025 and 2050 or 2030, it would suppress ultimate required natural uranium and control the uranium consumption about less than 1.2 million tons in 2150, while the FLWR in 2025 and FBR with breeding ratio of 1.16 in 2050 will at 0.9 million tons after in 2100. (T. Tanaka)

Analysis of fission gas release in LWR fuel using the BISON code

Energy Technology Data Exchange (ETDEWEB)

G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

2013-09-01

Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

Issues for Conceptual Design of AFCF and CFTC LWR Spent Fuel Separations Influencing Next-Generation Aqueous Fuel Reprocessing

International Nuclear Information System (INIS)

D. Hebditch; R. Henry; M. Goff; K. Pasamehmetoglu; D. Ostby

2007-01-01

In 2007, the U.S. Department of Energy (DOE) published the Global Nuclear Energy Partnership (GNEP) strategic plan, which aims to meet US and international energy, safeguards, fuel supply and environmental needs by harnessing national laboratory R and D, deployment by industry and use of international partnerships. Initially, two industry-led commercial scale facilities, an advanced burner reactor (ABR) and a consolidated fuel treatment center (CFTC), and one developmental facility, an advanced fuel cycle facility (AFCF) are proposed. The national laboratories will lead the AFCF to provide an internationally recognized R and D center of excellence for developing transmutation fuels and targets and advancing fuel cycle reprocessing technology using aqueous and pyrochemical methods. The design drivers for AFCF and the CFTC LWR spent fuel separations are expected to impact on and partly reflect those for industry, which is engaging with DOE in studies for CFTC and ABR through the recent GNEP funding opportunity announcement (FOA). The paper summarizes the state-of-the-art of aqueous reprocessing, gives an assessment of engineering drivers for U.S. aqueous processing facilities, examines historic plant capital costs and provides conclusions with a view to influencing design of next-generation fuel reprocessing plants

Accelerator driven light water fast reactor (revisiting to the accelerator LWR fuel regenerator)

International Nuclear Information System (INIS)

Takahashi, H.; Zhang, J.

1999-01-01

A tight-latticed, high-enriched Pu fuel reactor cooled by water or by super-critical steam has a high neutron economy, similar to that of Na-or Pb-cooled fast reactor. Operating in a subcritical condition by providing spallation neutrons, this Pu-fueled reactor can run safely, despite the positive coolant void coefficients. It can be used to transmute the proliferation-prone Pu into proliferation-resistive U-233 fuel using thorium as the fertile material. Rather than employing the large linear accelerator proposed for the LWR fuel regenerator studied in the INFCE program, a small circular accelerator, such as a cyclotron or a Fixed Field Alternating Gradient Synchrotron (FFAG), can run a large power reactor in a slightly subcritical reactor using control rods, on-line fuel reshuffling, and slightly graded proton-beam injection. Some thoughts on improving the reliability of the proton accelerator, on transmutation of the long-lived fission products of Tc-99, and I-129, and the future direction of the development of the fast reactor are discussed. (author)

Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

International Nuclear Information System (INIS)

Heaberlin, S.W.; Baker, D.A.

1976-06-01

As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,

Monitoring fuel consumption for reciprocating engines

International Nuclear Information System (INIS)

Zebelean, D.C.

1991-01-01

This paper reports that Northwest Pipeline Corp. has experienced substantial fuel cost savings through a program that provides continuous monitoring of predicted consumption against actual usage for reciprocating engines. With the continuous monitoring program, the company can gauge specific fuel consumption, flag inefficient engines and determine possible problems with computer software or hardware used to measure engine operations. The plan was initiated as part of an overall effort to reduce pipe line operating cost to remain market competitive by reducing cost of services to the customer. One of the factors in determining Northwest Pipeline's cost of service is the cost of fueling engines to transport natural gas. In 1990, Northwest consumed approximately 9,600,000 MMBtu in engine fuel. Fuel gas always has been accounted for and measured. However, the thermal efficiency of the fuel gas consumed was never quantified. Engineering management set a goal in 1989 to audit fuel consumption monthly on the basis of actual fuel consumed compared to the manufacturer's predicted consumption curves. The fuel consumption comparison between actual consumption with manufacturer's predicted consumption decreased 4% in six months after the first report was published. Unbalanced engines, faulty spark plugs, fuel valves, engines requiring overhauls, and even computer software and hardware problems were found, based on this report. Total decrease in the comparison of 4% was not all realized. Approximately 1.5% of the decrease was due to revising the manufacturer's fuel consumption curves to correctly predict Northwest's vintage of engine

Poolside inspection, repair and reconstitution of LWR fuel elements. Proceedings of a Technical Committee meeting

International Nuclear Information System (INIS)

1998-11-01

The Technical Committee Meeting on Poolside Inspection, repair and reconstruction of LWR Fuel Elements was organize by IAEA upon the recommendations of the International Working Group on Fuel performance Technology and held in Switzerland in October 1997. The purpose of the Meeting was to review the state of art in the area of poolside inspection, repair and reconstruction of light water fuel elements and to evaluate the progress achieved in this area since previous IAEA Meetings on the same topic in 1981 and 1984. The Meeting provided a forum on exchange of information between utilities, fuel designers and other authorities and specialists on a topic of current interest and real concern to industries in many Member States. The respective technologies are widely used or planned to be used in order to identify elementary major causes of fuel failure and to improve fuel utilization by repair and subsequent reuse of fuel elements. The Proceedings includes papers presented at the Meeting each described by a separate abstract

Recycle of LWR actinides to an IFR

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)

Review of tellurium release rates from LWR fuel elements under accident conditions

International Nuclear Information System (INIS)

Lorenz, R.A.; Beahm, E.C.; Wichner, R.P.

1983-01-01

Although fission product tellurium presents a potentially significant radiohazard, its release and transport in source-term experiments is frequently overlooked because it does not possess a readily measurable, gamma emission; moreover, a recent study emphasized noble gas, iodine and cesium release from LWR fuel elements because of the large data base that exists for these materials. Some new tests show that in some cases tellurium may be held up in core material to a greater degree than previously assumed - an observation that prompts a careful reappraisal of the existing tellurium-release data and its chemical foundation

A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Fujishiro, Toshio

1984-03-01

Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

Directory of Open Access Journals (Sweden)

WEON-JU KIM

2013-08-01

Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

Bus fuel consumption model

Energy Technology Data Exchange (ETDEWEB)

Zargari, S.A. [Iran Univ. of Science and Technology, Teheran (Iran, Islamic Republic of); Khan, A.M. [Carleton Univ., Ottawa, ON (Canada). Dept. of Civil and Environmental Engineering

2000-07-01

The interest in rapid bus transit has increased sharply with the realization that modern metropolitan areas rely on public transit to provide for strong economies and communities. As a prevention tool against traffic congestion, deteriorating air quality and rising greenhouse gas emissions, this study of bus fuel consumption was designed to assist in the planning and management of rapid bus transit. The Australian Road Research Board's (ARRB) Road Fuel Consumption Model was used as a starting point. The estimations required were realized with the help of Newtonian Mechanics. The four states of vehicular traffic were examined: acceleration, cruise, deceleration, and idle. The estimated total power required from the engine to overcome resistance forces, to run vehicle accessories and overcome internal engine friction was calculated. The data for the standard and articulated bus was obtained from OC Transpo in Ottawa. The study permitted the authors to conclude that the estimations for the parameters for power requirements and fuel consumption for heavy duty vehicles are appropriate. The methodology for the estimation of fuel consumption on the Transitway, which is part of the rapid bus transit system, proved adequate. In addition, the methodology was useful to estimate fuel savings resulting from demand management strategies with potential for modal shift. 9 refs., 6 tabs.

Influence of some fabrication parameters and operating conditions on the PCI failure occurrence in LWR fuel rods

International Nuclear Information System (INIS)

Bouffioux, P.

1980-01-01

In recent LWR designs, the fuel rod failures are induced by a chemically assisted mechanical process, i.e. stress corrosion cracking. The analytical approach towards the analysis of PCI-SCC failures is mainly based on the predictions of the COMETHE code. The failure criteria rely on the concept of a stress threshold together with fission product availability. In the present paper, the use of the COMETHE code to minimize PCI induced clad failure occurrences is illustrated by parametric studies to define acceptable fuel specifications and reactor operating conditions (steady and transient). (author)

Characterization and chemistry of fission products released from LWR fuel under accident conditions

International Nuclear Information System (INIS)

Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

1984-01-01

Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

International Nuclear Information System (INIS)

Lee, Geun Hyeong; Kim, Hee Reyoung

2014-01-01

LWR uses fuel as 235 U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile 233 U when 232 Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster

Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

Energy Technology Data Exchange (ETDEWEB)

Lee, Geun Hyeong; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

LWR uses fuel as {sup 235}U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile {sup 233}U when {sup 232}Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster.

Passenger car fuel consumption survey

Energy Technology Data Exchange (ETDEWEB)

1984-03-01

This survey originated from a proposal to monitor the fuel consumption and fuel economy of personal use passenger cars operated in Canada. Its purpose is to establish a data base which would contain information on total distance travelled, total amount of fuel consumed, average distance obtained per unit of fuel, total expenditures on fuel, and seasonal fluctuations in fuel consumption and in distance travelled. Among the needs served by this data base are the monitoring of passenger car fuel economy standards and the estimation of pasenger car fuel requirements in conditions involving fuel shortages. Survey methodology is by telephone interview to trace selected vehicles to the registered owners, at which time a fuel purchase diary is then mailed to the principal driver of the car. The results are tabulated on a quarterly basis and to be released as they become available in bulletins similar to this. Data are presented for each province and the total for Canada is given. During the fourth quarter of 1982, it is estimated that there were 7.3 million personal use passenger cars operated in Canada. These cars were driven 28 billion kilometers and consumed 4.3 billion litres of fuel. Their average litres/100 kilometres and the average fuel consumption was 590 litres. 8 tabs.

Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

International Nuclear Information System (INIS)

Jenks, G.H.

1979-12-01

Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

Implications of plutonium utilization strategies on the transition from a LWR economy to a breeder economy

International Nuclear Information System (INIS)

Newman, D.F.; Fleischman, R.M.; White, M.K.

1977-02-01

The plutonium interface between the LWR and LMFBR fuel cycles is examined for typical nuclear growth projections both with and without plutonium recycle in LWRs. In order to guarantee a fuel supply for projected LMFBR growth rates, significant multiple Pu recycle in LWRs will not be possible. However, about 78% of the benefit of multiple plutonium recycle between now and the turn of the century is realized by one recycle and then stockpiling spent MOX for the LMFBR. LMFBR reprocessing schecules are estimated based on accumulation of reprocessing load. These schedules are used to estimate the amount of plutonium recovered from LMFBR fuels and determine the residual LWR plutonium required to meet LMFBR demand. The stockpile of LWR produced plutonium in spent MOX is sufficient to fuel the LMFBR until commercial LMFBR reprocessing can be justified. After that time, recycle of plutonium in LWRs will be significantly limited by a continuing LMFBR demand for LWR plutonium due to the projected high LMFBR growth rate. LWR reprocessing requirements are estimated for the assumed condition that LWR plutonium recycle is not approved, but the LMFBR is still pursued as an energy option. The uncertainties presented by this condition are addressed qualitatively. However, in our judgment these uncertainties in the plutonium market would likely delay LMFBR growth to levels significantly below current projections

A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

International Nuclear Information System (INIS)

Kim, Chang Hyun

1997-02-01

A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

ALMR potential for actinide consumption

International Nuclear Information System (INIS)

Cockey, C.L.; Thompson, M.L.

1992-01-01

The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) sponsored fast reactor design based on the Power Reactor, Innovative Small Module (PRISM) concept originated by General Electric. This reactor combines a high degree of passive safety characteristics with a high level of modularity and factory fabrication to achieve attractive economics. The current reference design is a 471 MWt modular reactor fueled with ternary metal fuel. This paper discusses actinide transmutation core designs that fit the design envelope of the ALMR and utilize spent LWR fuel as startup material and for makeup. Actinide transmutation may be accomplished in the ALMR core by using either a breeding or burning configuration. Lifetime actinide mass consumption is calculated as well as changes in consumption behavior throughout the lifetime of the reactor. Impacts on system operational and safety performance are evaluated in a preliminary fashion. Waste disposal impacts are discussed. (author)

Electronuclear fissile fuel production. Linear accelerator fuel regenerator and producer LAFR and LAFP

International Nuclear Information System (INIS)

Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

1978-04-01

A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for 3 to 1000 MW(e) LWR power reactors over its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center

Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

Energy Technology Data Exchange (ETDEWEB)

Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

2008-07-01

The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

Status of the development of RU-43 fuel at INR Pitesti

International Nuclear Information System (INIS)

Horhoianu, G.

2008-01-01

More than 50000 fuel bundles containing natural uranium fuel have been irradiated in the CANDU-6 reactors of Cernavoda-Romania NPP, with a very low defect rate, to a core-average discharge burnup of 170-190 Mwh/kgU. Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.1% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improves fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in Cernavoda NPP. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper. (orig.)

Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

Energy Technology Data Exchange (ETDEWEB)

Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

2006-08-15

Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base

International Nuclear Information System (INIS)

Jones, K.E.; Moore, R.S.

1990-08-01

The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184

Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

International Nuclear Information System (INIS)

Peehs, M.; Garzarolli, F.; Goll, W.

1999-01-01

Although the safety of a dry long-term spent fuel store is scarcely influenced if a few fuel rods start to leak during extended storage - since all confinement systems are designed to retain gaseous activity safely - it is a very conservative safety goal to avoid the occurrence of systematic rod defects. To assess the extended storage performance of a spent fuel assembly (FA), the experience can be collated into 3 storage modes: I - fast rate of temperature decrease δ max ≥ δ ≥ 300 deg. C, II - medium rate of decrease for the fuel rod dry storage temperature 300 deg. C > δ ≥ 200 deg. C, III - slow to negligible rate of temperature decrease for δ 2 -fuel are practically immobile during storage. Consequently all fission-product-driven defect mechanisms will not take place. The leading defect mechanism - also for fuel rods with increased burnup - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. The allowable uniform strain of the cladding is 1 - 2%. Calculations were performed to predict the dry storage performance of fuel assemblies with a burnup ≤ 55 GW · d/tHM based on the fuel assemblies end of life (EOL)-data and on a representative curve T = f(t). The maximum allowable hot spot temperature of a fuel rod in the CASTOR V cask was between 348 deg. C (U FA) and 358 deg. C (MOX FA). The highest hoop strain predicted after 40 years of storage is 0.77% proving that spent LWR fuel dry storage is safe. (author)

Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

International Nuclear Information System (INIS)

Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

2009-01-01

The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

Vehicle technologies, fuel-economy policies, and fuel-consumption rates of Chinese vehicles

International Nuclear Information System (INIS)

Huo Hong; He Kebin; Wang, Michael; Yao Zhiliang

2012-01-01

One of the principal ways to reduce transport-related energy use is to reduce fuel-consumption rates of motor vehicles (usually measured in liters of fuel per 100 km). Since 2004, China has implemented policies to improve vehicle technologies and lower the fuel-consumption rates of individual vehicles. Policy evaluation requires accurate and adequate information on vehicle fuel-consumption rates. However, such information, especially for Chinese vehicles under real-world operating conditions, is rarely available from official sources in China. For each vehicle type we first review the vehicle technologies and fuel-economy policies currently in place in China and their impacts. We then derive real-world (or on-road) fuel-consumption rates on the basis of information collected from various sources. We estimate that the real-world fuel-consumption rates of vehicles in China sold in 2009 are 9 L/100 km for light-duty passenger vehicles, 11.4 L/100 km for light-duty trucks, 22 L/100 km for inter-city transport buses, 40 L/100 km for urban transit buses, and 24.9 L/100 km for heavy-duty trucks. These results aid in understanding the levels of fuel consumption of existing Chinese vehicle fleets and the effectiveness of policies in reducing on-road fuel consumption, which can help in designing and evaluating future vehicle energy-efficiency policies. - Highlights: ► Vehicle fuel-consumption rate (VFCR) data are rarely available in China. ► We review the fuel-economy policies currently in place in China and their impacts. ► We derive real-world VFCRs on the basis of information collected from various sources. ► Results aid in understanding the fuel consumption levels of Chinese vehicle fleets. ► Results help in designing and evaluating future vehicle energy-efficiency policies.

Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

International Nuclear Information System (INIS)

Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

1996-01-01

Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

The transuranic mass balance during the introduction of metal fuel FBR cycle

International Nuclear Information System (INIS)

Yokoo, Takeshi; Inoue, Tadashi

1999-01-01

The mass flow of plutonium and minor actinides is calculated for a future light water reactor-fast breeder reactor (LWR-FBR) transition scenario, in which power generation by LWRs is continued on a certain scale for a long period before the replacement by FBRs begins. The burnup of the LWR spent fuel is considered to be higher than the current standard. It is assumed that all the plutonium and minor actinides recovered from LWRs are used to start up and feed metal fuel commercial FBRs, which replace those LWRs that have reached the end of their life. The results show that the accumulated plutonium and minor actinides from the LWRs can be consistently consumed without further accumulation, by gradually establishing the FBR power generation and its fuel cycle on the same scale. The optimum content of the minor actinides in the standard FBR fuel is about 2 weight percents. This result indicates that if FBRs are introduced in the future, extension of the LWR usage period will cause no significant problems in terms of the consumption of accumulated transuranic elements. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-01

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

LWR and HTGR coolant dynamics: the containment of severe accidents

International Nuclear Information System (INIS)

Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

1983-07-01

This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

Wood fuels utilization in Central Europe - the wood fuels consumption and the targets of utilization

International Nuclear Information System (INIS)

Alakangas, E.

1999-01-01

Following subjects are discussed in this presentation: The share of bioenergy of the total energy consumption in EU region; the wood fuels consumption in EU region in 1995; the division of bioenergy utilization (households, wood- based district heating, wood consumption in industry, power generation from wood and residues, biofuels, biogas and sludges); wood fuels consumption in households in EU countries in 1995; wood consumption in France; the additional wood fuel consumption potential in France; Blan bois - wood energy program; French wood energy markets; German wood energy markets; energy consumption in Germany; wood consumption in Bavaria; the wood fuels potential in Bavaria; wood fuels consumption in households in Bavaria; wood fuels consumption for district heating in Bavaria; fuel prices in Bavaria; Environmental regulations in Germany; small boiler markets in Germany; Energy consumption in Austria; small-scale utilization of wood fuels; utilization of wood energy. (Slides, additional information from the author)

PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

International Nuclear Information System (INIS)

Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

2001-01-01

A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

Experience of European LWR irradiated fuel transport: the first five hundred tonnes

International Nuclear Information System (INIS)

Curtis, H.W.

1978-01-01

The paper describes the service provided by an international company specializing in the transport of LWR irradiated fuel throughout Europe. Methods of transport used to the reprocessing plants at La Hague and Windscale include road transport of 38 te flasks over the whole route; transport of flasks between 55 and 105 te by rail, with rail-head and the reprocessing plant, where required, performed by road using heavy trailers; roll-on, roll-off sea ferries; and charter ships. Different modes of transport have been developed to cater for the various limitations on access to reactor sites arising from geographical and routing considerations. The experience of transporting more than 500 tonnes of irradiated uranium from twenty-one power reactors is used to illustrate the flexibility which the transport organization requires when the access and handling facilities are different at almost every reactor. Variations in fuel cross sections and lengths of fuel elements used in first generation reactors created the need for first generation flasks with sufficient variants to accommodate all reactor fuels but the trend now is towards standardization of flasks to perhaps two basic types. The safety record of irradiated fuel transport is examined with explanation of the means whereby this has been achieved. The problems of programming the movement of a pool of eighteen flasks for twenty-one reactors in eight countries are discussed together with the steps taken to ensure that the service operates fairly to give priority to those reactors with the greatest problems. The transport of irradiated fuel across several national frontiers is an international task which requires an international company. The transport of European irradiated fuel can be presented as an example of international collaboration which works

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

International Nuclear Information System (INIS)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

2013-01-01

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

Energy Technology Data Exchange (ETDEWEB)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

2013-07-01

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

Fuel consumption impacts of auto roof racks

International Nuclear Information System (INIS)

Chen, Yuche; Meier, Alan

2016-01-01

The after-market roof rack is one of the most common components attached to a vehicle for carrying over-sized items, such as bicycles and skis. It is important to understand these racks’ fuel consumption impacts on both individual vehicles and the national fleet because they are widely used. We estimate the national fuel consumption impacts of roof racks using a bottom-up approach. Our model incorporates real-world data and vehicle stock information to enable assessing fuel consumption impacts for several categories of vehicles, rack configurations, and usage conditions. In addition, the model draws on two new data-gathering techniques, on-line forums and crowd-sourcing. The results show that nationwide, roof racks are responsible for 0.8‰ of light duty vehicle fuel consumption in 2015, corresponding to 100 million gallons of gasoline per year. Sensitivity analyses show that results are most sensitive to the fraction of vehicles with installed roof racks but carrying no equipment. The aerodynamic efficiency of typical roof racks can be greatly improved and reduce individual vehicle fuel consumption; however, government policies to minimize extensive driving with empty racks—if successful—could save more fuel nationally. - Highlights: •First estimate of national energy impacts of auto roof racks—about 1‰. •A bottom-up approach reveals details of the fuel consumption penalty caused by racks. •Two novel data collection techniques, on-line forums and crowd-sourcing, improve estimate. •Technical and behavioral policies could significantly cut fuel penalties from roof racks.

Determinants of fuel consumption in mining trucks

International Nuclear Information System (INIS)

Dindarloo, Saeid R.; Siami-Irdemoosa, Elnaz

2016-01-01

Analysis of fuel consumption in a large surface mine, during more than 5000 cycles of material transportation, revealed considerable variability in the data. Truck fuel estimation based on the mining truck manufacturers' manuals/estimates is not capable of capturing this variability in the fuel consumption data. Partial least squares regression and autoregressive integrated moving average methods were employed to examine the effect of cyclic activities on fuel consumption, and to recommend relevant remedies for consumption reduction. Proper modifications of the operation can result in improved cycle times. Consequently, minimizing some cyclic activities would enhance energy efficiency. The truck “empty idle time” was a major contributor to unnecessary fuel consumption. Since the truck queues at shovels are a major component of the “empty idle time”, decisions should be reviewed to reduce the truck queues at loading points. Improved dispatching strategies, optimal muck pile shape and size distribution, and improved shovel/loader operator skills are effective preventive measures to minimize truck flow bottlenecks at loading points, and thus to improve energy efficiency at mines. - Highlights: • A large fleet of mining trucks consumes considerable amounts of energy. • Current energy models do not consider the effect of site-specific mining conditions. • A new methodology based on material handling cyclic activities is proposed. • Fuel consumption rates in different truck operating modes are determined. • The new model is compared with the truck manufacturer's energy consumption guideline.

Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

International Nuclear Information System (INIS)

Cerba, S.; Hrncir, M.; Necas, V.

2012-01-01

The study deals with the verification of the depletion capabilities of the MCNPX code, which is a linked Monte-Carlo depletion code. For such a purpose the IV-B phase of the OECD NEA Burnup credit benchmark has been chosen. The mentioned benchmark is a code to code comparison of the multiplication coefficient k eff and the isotopic composition of a LWR MOX fuel assembly at three given burnup levels and after five years of cooling. The benchmark consists of 6 cases, 2 different Pu vectors and 3 geometry models, however in this study only the fuel assembly calculations with two Pu vectors were performed. The aim of this study was to compare the obtained result with data from the participants of the OECD NEA Burnup Credit project and confirm the burnup capability of the MCNPX code. (Authors)

Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

International Nuclear Information System (INIS)

Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

2014-01-01

An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods

International Nuclear Information System (INIS)

Lemehov, Sergei; Motoe, Suzuki

2003-01-01

1 - Description of program or function: The PLUTON-PC is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO 2 , UO 2 -Gd 2 O 3 , inhomogeneous MOX, and UO 2 -ThO 2 . The PLUTON-PC code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. 2 - Methods: Based upon cumulative yields, the PLUTON-PC code calculates as a function of radial position and local burnup concentrations of fission products, macroscopic scattering cross-sections and self-shielding effect which is important for standard fuel (for Pu-242 mainly) and more importantly for homogeneous and inhomogeneous MOX fuel because of higher concentrations of fissile and fertile isotopes of plutonium. The code results in burnup dependent fission rate density profiles throughout the in-reactor irradiation of LWR fuel rods. The isotopes included in calculations have been extended to cover all trans-uranium groups (plutonium plus higher actinides) of fissile and fertile isotopes. Self-shielding problem and scattering effects have been revised and solved for all isotopes in the calculations for adequacy at high burnup, different irradiation conditions and cladding materials

Air quality impacts due to construction of LWR waste management facilities

International Nuclear Information System (INIS)

1977-06-01

Air quality impacts of construction activities and induced housing growth as a result of construction activities were evaluated for four possible facilities in the LWR fuel cycle: a fuel reprocessing facility, fuel storage facility, fuel fabrication plant, and a nuclear power plant. Since the fuel reprocessing facility would require the largest labor force, the impacts of construction of that facility were evaluated in detail

Fuels. Deliveries and consumption of fuels during 4th quarter 2002 and year 2002

International Nuclear Information System (INIS)

2003-01-01

The total volume of oil product deliveries (excluded lubricants, bitumen etc.) from the oil companies was 1 per cent lower during 2002 compared with 2001. The deliveries of motor gasoline and diesel oil increased by 2 and 5 per cent. Domestic heating oil and heavy fuel oils decreased by 3 and 5 per cent compared with 2001. Under the 4th quarter 2002 the deliveries of oil products (excluded lubricants, bitumen etc.) was 10 per cent higher compared with the same quarter 2001. The deliveries of domestic heating oil and heavy fuel oils increased by 22 per cent in each case. Motor gasoline was unchanged while the deliveries of diesel oil increased by 7 per cent. The consumption of fuels in mining and manufacturing increased by 3 % during the year 2002. The consumption of hard coal and coke increased by 5 % and the rest of the fossil fuels by 7 %. Renewable fuels decreased by 1 %. During the 4th quarter 2002 increased the fuel consumption in industry by 2 % compared with same quarter 2001. Renewable fuels decreased by 7 % while fossil fuels increased by 18 %. The consumption of fuels in electricity, gas and district heating services increased by 12 % during 2002. The consumption of oils increased by 16 %. During year 2002 the renewable fuels accounts for 68 % of the final consumption. During the 4th quarter the consumption of fuels was almost twice as high as the consumption during the 4th quarter 2001

Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

1978-09-01

In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

Conceptual study of the future nuclear fuel cycle system for the extended LWR age

International Nuclear Information System (INIS)

Fujine, Sachio; Takano, Hideki; Sato, Osamu; Tone, Tatsuzo; Yamada, Takashi; Kurosawa, Katsutoshi.

1993-08-01

A large scale integrated fuel cycle facility (IFCF) is assumed for the future nuclear fuel cycle in the extended LWR age. Spent MOX fuels are reprocessed mixed with UOX in a centralized reprocessing plant. The reprocessing plant separates long-lived nuclides as well as Pu. Nitric acid solutions of those products are fed directly to MOX fabrication process which is incorporated with reprocessing. MOX pellets are made by sphere-cal process. Two process concepts are made as advanced reprocessing incorporated with partitioning (ARP) which has the function of long-lived nuclides recovery. One is a simplified Purex combined with partitioning. Extractable long-lived nuclides, 237 Np and 99 Tc, are assumed to be recovered in main flow stream of the improved Purex process. The other process concept is made aiming at recovering all TRU nuclides in reprocessing to meet with TRU recycle requirement in the long future. A concept of the future fuel cycle system is made by combining integrated fuel cycle facility and very high burnup LWRs (VHBR). The reactor concept of VHBRs has been proposed to improve Pu recycle economy in the future. Highly enriched MOX fuel are loaded in the full core of reactor in order to increase reactivity for the burnup. Fuel cycle indices such as Pu isotopic composition change, spent fuel integration, nuclide transmutation effect are estimated by simulating the Pu recycling in the system of VHBR and ARP. It is concluded that Pu enrichment of MOX fuel can be kept less than 20 % through multi-recycle. Reprocessing MOX fuels with UOX shows a favorable effect for keeping Pu reactivity high enough for VHBR. Integration of spent MOX fuel can be reduced by Pu recycle. Transmutation of Np is feasible by containing Np into MOX fuel. (author)

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

International Nuclear Information System (INIS)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

2003-01-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

International Nuclear Information System (INIS)

Heuser, Brent

2013-01-01

An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

Estimating road transport fuel consumption in Ecuador

International Nuclear Information System (INIS)

Sierra, Jaime Cevallos

2016-01-01

Road transport is one of the sectors with highest energy consumptions in the planet, with large dependence of fossil fuels, and contribution for global greenhouse gas emissions. Although, Latin America is not a high-energy consumer, its share in global consumption is expected to grow, especially in the transportation sector. This make essential for developing countries the adoption of better policies to identify the vehicle groups with largest fuel demands. The present study describes the VKT technique to disaggregate road transport energy consumption by vehicle type, applied to the road transportation system of Ecuador. It also describes the procedures performed to estimate the variables required to run the model, and some of the practical applications that be used to create public policies. Results show as the biggest fuel consumers the heavy-duty freight cargo, followed by light duty vehicles. The estimation of greenhouse gas emissions evidence that road transport released 14.3 million tons of CO_2 in 2012. When fuel consumption is compared by it costs, it can be confirmed that Ecuadorean Government covered, through subsidies, for 68% of the annual fuel costs of national road transport, demonstrating the importance of restructuring these expenditures in order to achieve an efficient road transport system. - Highlights: •The vehicle-kilometers traveled has been estimated from local info. •The fuel economy has been calculated from national and international data. •The groups with higher fuel consumption has been located. •The fuel-type dependency has been estimated for each vehicle group. •Greenhouse gas emission, and fuel costs, has been estimated for local road transport.

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-07-25

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

International Nuclear Information System (INIS)

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

2017-01-01

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

Light and Heavy Tactical Wheeled Vehicle Fuel Consumption Evaluations Using Fuel Efficient Gear Oils (FEGO)

Science.gov (United States)

2016-05-01

UNCLASSIFIED LIGHT AND HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FUEL EFFICIENT GEAR OILS (FEGO) FINAL... HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FUEL EFFICIENT GEAR OILS (FEGO) FINAL REPORT TFLRF No. 477 by Adam C...August 2014 – March 2016 4. TITLE AND SUBTITLE LIGHT AND HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FEUL EFFICIENT GEAR OILS

Compilation of papers presented to the KTG conference on 'Advanced LWR fuel elements: Design, performance and reprocessing', 17-18 November 1988, Karlsruhe Nuclear Research Center

International Nuclear Information System (INIS)

Bahm, W.

1989-05-01

The two expert groups of the Nuclear Society (KTG), 'chemistry and waste disposal' and 'fuel elements' discussed interdisciplinary problems concerning the development and reprocessing of advanced fuel elements. The 10 lectures deal with waste disposal, mechanical layout, operating behaviour, operating experiences and new developments of fuel elements for water moderated reactors as well as operational experiences of the Karlsruhe reprocessing plant (WAK) with reprocessing of high burnup LWR and MOX fuel elements, the distribution of fission products, the condition of the fission products during dissolution and with the effects of the higher burnup of fuel elements on the PUREX process. (DG) [de

An Optimization Study of LWR Fuel Assembly Design for TRU Burning using FCM and UO{sub 2}-ThO{sub 2} Fuel Pins

Energy Technology Data Exchange (ETDEWEB)

Hwang, Daehee; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

2014-05-15

The objective of this work is to design optimized LWR fuel assemblies for the transmutation of TRU (transuranic) nuclides by using FCM (Fully Ceramic Micro-encapsulated) and UO{sub 2}-ThO{sub 2} fuel pins without degradation of safety-related parameters. In our study, the pin pitch (equivalently to P/D (Pitch-to-Diameter) ratio with a fixed fuel rod diameter) is used as a design parameter. The motivation is to make MTC (Moderator Temperature Coefficient) less negative at EOC because it was found that the small LWR core design in our previous work has a very strong MTC at EOC (∼-80pcm/K) which can lead to a large positive reactivity insertion under MSLB (Main Steam Line Break) accident and to a reduction of shutdown margin of the control rods. The basic idea is to increase moderator-to-fuel ratio such that the fuel assemblies have less negative MTC due to increase the moderation. The results show that a small increase of P/D ratio by 3.8% can give a considerably less negative MTC and an increase of TRU destruction rate without an increase of pin power peaking. In our study, a special emphasis is given on the effects of the increased P/D ratio for MTC. From the results, it was found that an increase of P/D ratio (we considered up to P/D=1.38) leads to a less negative MTC and a less negative FTC, an increase of TRU destruction rate, and a decrease of {sup 233}U production in UO{sub 2}-ThO{sub 2} pins. In particular, a small change of P/D ratio from 1.33 to 1.38 led to a change of MTC from - 75 pcm/.deg. C to -67 pcm/.deg. C at EOC, and a small increase of net TRU destruction rate from 26.4% to 28.3%. As conclusion, a small increase of P/D ratio is effective in obtaining the less negative MTC at EOC with a small increase of TRU destruction rate and without a significant degradation of FTC.

Predicting vehicle fuel consumption patterns using floating vehicle data.

Science.gov (United States)

Du, Yiman; Wu, Jianping; Yang, Senyan; Zhou, Liutong

2017-09-01

The status of energy consumption and air pollution in China is serious. It is important to analyze and predict the different fuel consumption of various types of vehicles under different influence factors. In order to fully describe the relationship between fuel consumption and the impact factors, massive amounts of floating vehicle data were used. The fuel consumption pattern and congestion pattern based on large samples of historical floating vehicle data were explored, drivers' information and vehicles' parameters from different group classification were probed, and the average velocity and average fuel consumption in the temporal dimension and spatial dimension were analyzed respectively. The fuel consumption forecasting model was established by using a Back Propagation Neural Network. Part of the sample set was used to train the forecasting model and the remaining part of the sample set was used as input to the forecasting model. Copyright © 2017. Published by Elsevier B.V.

Introduction to nuclear supply chain management. In the context of fuel cycle strategy from LWR cycle system to FR cycle system

International Nuclear Information System (INIS)

Shiotani, Hiroki; Ono, Kiyoshi; Namba, Takashi; Yasumatsu, Naoto; Heta, Masanori

2011-01-01

Supply chain management (SCM) is an important technique to maintain supply and demand balance and to achieve total optimization from upstream to downstream in manufacturers' management. One of the major reasons why SCM receives much attention recently is the trend in production and sales systems from 'Push type' to 'Pull type'. 'Push type' can be restated as 'Make to Stock' (MTS). MTS is a type of supply chain in which the production is not connected to actual demand. On the contrary, 'Pull type' can be restated as 'Make to Order' (MTO) in which the production is connected to actual demand. In this paper, the terminologies and ideas of SCM was introduced into the scenario study to give a fresh perspective for considering LWR cycle to FR cycle transition strategies in Japan. Then, an analytical tool (SCM tool) which has been developed by the authors is used to survey Japanese nuclear energy system in transition with the SCM terminologies and viewpoints. When some of the Japanese nuclear fuel cycle strategies and tools are thought back with the framework of SCM, they tend to treat nuclear fuel cycle system as 'Push type' supply chain in their simulations. For example, a reprocessing plant separates SFs (spent fuels) without considering the actual Pu demand. However, because future reprocessing plants and fuel fabrication plants will act as Pu suppliers (front-end facility) to FR as well as back-end facilities of LWRs, the reasonable plant operation principle can be 'Pull type'. The analysis was conducted by the SCM tool to simulate the behaviors of both MTS and MTO type facilities during the LWR to FR transition period. If there are large uncertainties in the Pu demand or the load factor, etc. of future reprocessing plants, SCM framework is beneficial. Furthermore, the realization of MTO type operation by SCM can reduce the recovered Pu stock in spite of the increase of the SF interim storage. As the result of the investigation on the boundary location of 'Push type

Modelling the effects of transport policy levers on fuel efficiency and national fuel consumption

International Nuclear Information System (INIS)

Kirby, H.R.; Hutton, B.; McQuaid, R.W.; Napier Univ., Edinburgh; Raeside, R.; Napier Univ., Edinburgh; Zhang, Xiayoan; Napier Univ., Edinburgh

2000-01-01

The paper provides an overview of the main features of a Vehicle Market Model (VMM) which estimates changes to vehicle stock/kilometrage, fuel consumed and CO 2 emitted. It is disaggregated into four basic vehicle types. The model includes: the trends in fuel consumption of new cars, including the role of fuel price: a sub-model to estimate the fuel consumption of vehicles on roads characterised by user-defined driving cycle regimes; procedures that reflect distribution of traffic across different area/road types; and the ability to vary the speed (or driving cycle) from one year to another, or as a result of traffic growth. The most significant variable influencing fuel consumption of vehicles was consumption in the previous year, followed by dummy variables related to engine size. the time trend (a proxy for technological improvements), and then fuel price. Indeed the effect of fuel price on car fuel efficiency was observed to be insignificant (at the 95% level) in two of the three versions of the model, and the size of fuel price term was also the smallest. This suggests that the effectiveness of using fuel prices as a direct policy tool to reduce fuel consumption may he limited. Fuel prices may have significant indirect impacts (such as influencing people to purchase more fuel efficient cars and vehicle manufacturers to invest in developing fuel efficient technology) as may other factors such as the threat of legislation. (Author)

Fuel consumption modeling in support of ATM environmental decision-making

Science.gov (United States)

2009-07-01

The FAA has recently updated the airport terminal : area fuel consumption methods used in its environmental models. : These methods are based on fitting manufacturers fuel : consumption data to empirical equations. The new fuel : consumption metho...

Generalized perturbation theory for LWR depletion analysis and core design applications

International Nuclear Information System (INIS)

White, J.R.; Frank, B.R.

1986-01-01

A comprehensive time-dependent perturbation theory formulation that includes macroscopic depletion, thermal-hydraulic and poison feedback effects, and a criticality reset mechanism is developed. The methodology is compatible with most current LWR design codes. This new development allows GTP/DTP methods to be used quantitatively in a variety of realistic LWR physics applications that were not possible prior to this work. A GTP-based optimization technique for incore fuel management analyses is addressed as a promising application of the new formulation

A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

Energy Technology Data Exchange (ETDEWEB)

Blomeke, J.O.

2001-07-10

The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

Energy Technology Data Exchange (ETDEWEB)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

2003-07-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

Fuel consumption models for pine flatwoods fuel types in the southeastern United States

Science.gov (United States)

Clinton S. Wright

2013-01-01

Modeling fire effects, including terrestrial and atmospheric carbon fluxes and pollutant emissions during wildland fires, requires accurate predictions of fuel consumption. Empirical models were developed for predicting fuel consumption from fuel and environmental measurements on a series of operational prescribed fires in pine flatwoods ecosystems in the southeastern...

An examination of fuel consumption trends in construction projects

International Nuclear Information System (INIS)

Peters, Valerie A.; Manley, Dawn K.

2012-01-01

Recent estimates of fuel consumption in construction projects are highly variable. Lack of standards for reporting at both the equipment and project levels make it difficult to quantify the magnitude of fuel consumption and the associated opportunities for efficiency improvements in construction projects. In this study, we examined clusters of Environmental Impact Reports for seemingly similar construction projects in California. We observed that construction projects are not characterized consistently by task or equipment. We found wide variations in estimates for fuel use in terms of tasks, equipment, and overall projects, which may be attributed in part to inconsistencies in methodology and parameter ranges. Our analysis suggests that standardizing fuel consumption reporting and estimation methodologies for construction projects would enable quantification of opportunities for efficiency improvements at both the equipment and project levels. With increasing emphasis on reducing fossil fuel consumption, it will be important to quantify opportunities to increase fuel efficiency, including across the construction sector. - Highlights: ► An analysis of construction projects reveals inconsistencies in fuel use estimates. ► Fuel consumption estimates for similar construction equipment can vary greatly. ► Standards would help to quantify efficiency opportunities in construction.

Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

International Nuclear Information System (INIS)

DelCul, Guillermo D.; Trowbridge, Lee D.; Renier, John-Paul; Ellis, Ronald James; Williams, Kent Alan; Spencer, Barry B.; Collins, Emory D.

2009-01-01

This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the 235 U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of 238 Pu due to the presence of 236 U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance

Stepwise evolution of fuel assembly design toward a sustainable fuel cycle with hard neutron spectrum light water reactors

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro

2011-01-01

An advanced LWR with hard neutron spectrum, FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design keeping the same core configuration. A proposed evolution process of the design toward a sustainable fuel cycle is composed of three stages, the first one based on the LWR fuel cycle infrastructures, the second one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the third one based on the FR fuel cycle infrastructures. For the first stage, a fuel assembly design concept named FLWR/MIX has been developed in which enriched UO 2 fuel rods are arranged in the peripheral region of the assembly, surrounding the MOX fuel rods in the central region. The FLWR/MIX design realizes a breeder type operation under the framework of the LWR-MOX technologies and there experience. A modified FLWR/MIX design with low Pu inventory for the second stage has a potential of high Puf conversion ratio of 1.1 and can contribute to smooth and speedy transition from the LWR fuel cycle to the FR fuel cycle. For the third stage, the FLWR/MIX design is extended into a design with natural UO 2 fuel rods to realize multiple Pu recycling keeping a Puf conversion ratio of around 1.0. (author)

Fuels. Deliveries and consumption of fuels during 4th quarter 2001 and during 2001

International Nuclear Information System (INIS)

2002-01-01

The total volume of oil product deliveries (excluded lubricants, bitumen etc.) from the oil companies was 3 per cent higher during 2001 compared with 2000. The deliveries of motor gasoline and diesel oil increased by 1 per cent. Domestic heating oil decreased by 2 per cent and heavy fuel oils increased by 17 per cent, all compared with 2000. Under the forth quarter 2001 the deliveries of oil products (excluded lubricants bitumen etc.) was 5 per cent higher compared with the same quarter 2000. The deliveries of domestic heating oil and heavy fuel oils increased by 4 per cent and 28 per cent respectively. Motor gasoline increased by 3 per cent while the deliveries of diesel oil decreased by 1 per cent. The consumption of fuels in mining and manufacturing decreased by 5 per cent during 2001 compared with 2000. Most of the reduction came from the lower use of domestic fuel in manufacture of pulp, paper and paper products, printing and publishing. The total consumption in mining and manufacturing was unchanged during the 4 th quarter 2001 compared with the same period 2000. The consumption of domestic fuels decreased by 6 per cent and the oil consumption increased by 11 per cent. Higher consumption of fuels in electricity, gas and district heating services. The consumption of fuels in electricity, gas and district heating services increased by 14 per cent 2001 compared with 2000. The consumption of domestic fuels raised with 12 per cent and the use of fossil fuels raised with 17 per cent. The highest increasing, 27 per cent, was found within the use of oil. Electricity, gas and district heating services increased there fuel consumption with 29 per cent during the 0 quarter 2001 compared with the same period 2000. The raise can be explained with the fact that the 4th quarter 2000 was unusually warm. The increasing is found both in the use of domestic and fossil fuels

LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

Energy Technology Data Exchange (ETDEWEB)

Peehs, M [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-12-01

Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves.

LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

International Nuclear Information System (INIS)

Peehs, M.

1997-01-01

Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves

Fuel consumption and emission on fuel mixer low-grade bioethanol fuelled motorcycle

Directory of Open Access Journals (Sweden)

Abikusna Setia

2017-01-01

Full Text Available Bioethanol is currently used as an alternative fuel for gasoline substitute (fossil fuel because it can reduce the dependence on fossil fuel and also emissions produced by fossil fuel which are CO2, HO, NOx. Bioethanol is usually used as a fuel mixed with gasoline with certain comparison. In Indonesia, the usage is still rare. Bioethanol that is commonly used is bioethanol anhydrous 99.5%. In the previous studies, bioethanol was distilled from low to high grade to produce ethanol anhydrous. But the result is only able to reach 95% or ethanol hydrous. This study is objected to design a simple mechanism in the mixing of bioethanol hydrous with the gasoline using a fuel mixer mechanism. By this mechanism, the fuel consumption and the resulting emissions from combustion engine can be analyzed. The fuel blend composition is prepared as E5, E10, and E15/E20, the result of fuel consumption and emission will be compared with pure gasoline. The using of bioethanol hydrous as a fuel mixture was tended to produce more stable bioethanol fuel consumption. However, the utilization of the mixture was found able to reduce the exhaust emissions (CO, HC, and NOx.

Development of uranium reduction system for incineration residue generated at LWR nuclear fuel fabrication plants in Japan

International Nuclear Information System (INIS)

Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.

1993-01-01

The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests

Estimating Vehicle Fuel Consumption and Emissions Using GPS Big Data

Directory of Open Access Journals (Sweden)

Zihan Kan

2018-03-01

Full Text Available The energy consumption and emissions from vehicles adversely affect human health and urban sustainability. Analysis of GPS big data collected from vehicles can provide useful insights about the quantity and distribution of such energy consumption and emissions. Previous studies, which estimated fuel consumption/emissions from traffic based on GPS sampled data, have not sufficiently considered vehicle activities and may have led to erroneous estimations. By adopting the analytical construct of the space-time path in time geography, this study proposes methods that more accurately estimate and visualize vehicle energy consumption/emissions based on analysis of vehicles’ mobile activities (MA and stationary activities (SA. First, we build space-time paths of individual vehicles, extract moving parameters, and identify MA and SA from each space-time path segment (STPS. Then we present an N-Dimensional framework for estimating and visualizing fuel consumption/emissions. For each STPS, fuel consumption, hot emissions, and cold start emissions are estimated based on activity type, i.e., MA, SA with engine-on and SA with engine-off. In the case study, fuel consumption and emissions of a single vehicle and a road network are estimated and visualized with GPS data. The estimation accuracy of the proposed approach is 88.6%. We also analyze the types of activities that produced fuel consumption on each road segment to explore the patterns and mechanisms of fuel consumption in the study area. The results not only show the effectiveness of the proposed approaches in estimating fuel consumption/emissions but also indicate their advantages for uncovering the relationships between fuel consumption and vehicles’ activities in road networks.

Estimating Vehicle Fuel Consumption and Emissions Using GPS Big Data.

Science.gov (United States)

Kan, Zihan; Tang, Luliang; Kwan, Mei-Po; Zhang, Xia

2018-03-21

The energy consumption and emissions from vehicles adversely affect human health and urban sustainability. Analysis of GPS big data collected from vehicles can provide useful insights about the quantity and distribution of such energy consumption and emissions. Previous studies, which estimated fuel consumption/emissions from traffic based on GPS sampled data, have not sufficiently considered vehicle activities and may have led to erroneous estimations. By adopting the analytical construct of the space-time path in time geography, this study proposes methods that more accurately estimate and visualize vehicle energy consumption/emissions based on analysis of vehicles' mobile activities ( MA ) and stationary activities ( SA ). First, we build space-time paths of individual vehicles, extract moving parameters, and identify MA and SA from each space-time path segment (STPS). Then we present an N-Dimensional framework for estimating and visualizing fuel consumption/emissions. For each STPS, fuel consumption, hot emissions, and cold start emissions are estimated based on activity type, i.e., MA , SA with engine-on and SA with engine-off. In the case study, fuel consumption and emissions of a single vehicle and a road network are estimated and visualized with GPS data. The estimation accuracy of the proposed approach is 88.6%. We also analyze the types of activities that produced fuel consumption on each road segment to explore the patterns and mechanisms of fuel consumption in the study area. The results not only show the effectiveness of the proposed approaches in estimating fuel consumption/emissions but also indicate their advantages for uncovering the relationships between fuel consumption and vehicles' activities in road networks.

Study on reprocessing plant during transition period from LWR to FBR

International Nuclear Information System (INIS)

Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

2011-01-01

We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

Wood fuel in Sweden 1800-1990 - consumption and price trends

International Nuclear Information System (INIS)

Schoen, L.

1992-01-01

The report presents consumption and price trends of wood fuel in Sweden 1800-1990 and discusses the increase in the use of wood fuel in the 1980's in a long-term perspective. Consumption of wood fuel grew at the same rate as population during most of the 19th century with a share of 95-80 per cent of total fuel consumption. Since the modern industrial breakthrough around the 1880's, consumption of wood fuel has decreased while that of fossiles and electricity have expanded. Temporarily, consumption increased during the world wars, particularly during the second one. The increase after the energy crises of the 1970's differs from those of the wars in some respects - thus, the changes in the conditions of energy supply and energy use were conceived as long-lasting, the increase in consumption took place with markets in function, and an important new user appeared, namely the district heating services. During both the 19th and the 20th century, prices of wood fuel have risen strongly in relation to those of most other products. This increase expresses the shifts in demand to wood resources and the comparatively weak productivity growth in forestry. Compared to prices of fossile fuels, the price increase of wood fuel ended in the 1920's and the relation has since then fluctuated. The strong shift in consumption to fossils from the 1920's is explained rather by the high costs of handling wood fuel. Wood fuel consumption has increased during periods of relatively decreasing wages. While the price increase of wood fuel can stimulate extended production also within agriculture, the study emphasizes the need of productivity growth in wood fuel production as well as product development towards the lowering of the handling costs of the user. (23 refs., 8 figs., 4 tabs.)

A comparative study on recycling spent fuels in gas-cooled fast reactors

International Nuclear Information System (INIS)

Choi, Hangbok; Baxter, Alan

2010-01-01

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

Consideration on the partial moderation in criticality safety analysis of LWR fresh fuel storage

International Nuclear Information System (INIS)

Tanaka, S.; Tanimoto, R.; Suzuki, K.; Ishitobi, M.

1987-01-01

In criticality safety analyses of fuel fabrication facilities, neutron effective multiplication factor (k eff ) of a storage vault has been calculated assuming ''partial moderation'' in whole space (hereafter reffered to as unlimited partial moderation). Where the enrichment of fuels to be stored is about 3.5 % or less, calculated k eff is usually low enough to show subcriticality even in unlimited partial moderation. However, it is scheduled to elevate LWR fuels enrichment for economical higher burnup and the unlimited partial moderation would require to introduce neutron absorbers to maintain subcriticality. It is clear that this causes economical disadvantages, and hence we reconsidered this assumption to avoid such a condition. Reconsideration of the unlimited partial moderation was carried out in following steps. (1) Water quantity to be assumed in atmosphere to obtain criticality was revealed too much to realize. (2) Typical realistic water quantity in atmosphere was estimated to apply as an alternative assumption. (3) A fresh fuel assembly storage was chosen as a model array and calculations with lattice code WIMS-D 1 and Monte Calro code KENO-IV 2 were performed to compare new alternative assumption with the unlimited one. As results of the above calculations, maximum k eff of the array under the new assumption was remarkably reduced to the value less than 0.95 though the maximum k eff under the unlimited one was higher than 1.0. (author)

Parameters' influence estimation on Puf supply and demand in transitional period from LWR to FBR in Japan

International Nuclear Information System (INIS)

Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

2009-01-01

Plutonium fissile (Puf) amounts to balance supply and demand during transition period were evaluated with different parameters. Estimated total Puf demand in transitional period was sensitive to deployment speed of FBR. Because FBRs will be deployed as replacements of old LWRs for keeping total capacity, deployment history of existing LWRs should be taken into consideration. According to the estimation, LWR fuel burnup and utilized capacity are not big issue. Because certain amount of LWR spent fuel will remain in early phase of transitional period, there is enough time for preparing Puf supply. On the other hand, FBR fuel cycle time (SF cooling time + fuel fabrication time) have large impact on Puf supply. Fuel cycle technologies including transportation for applying to short cooling spent fuels should be developed. (author)

Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

International Nuclear Information System (INIS)

Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

1978-01-01

The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

International Nuclear Information System (INIS)

Spino, J.; Peerani, P.

2008-01-01

The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release

Processing of the GALILEO fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology

International Nuclear Information System (INIS)

Mailhe, P.; Barbier, B.; Garnier, C.; Landskron, H.; Sedlacek, R.; Arimescu, I.; Smith, M.; Bellanger, P.

2013-01-01

The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEO along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEO code benchmarking process, on its extended experimental database and on the GALILEO model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEO model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples. With the submittal of Topical Report GALILEO to the U.S. NRC in 2013, GALILEO and its methodology are on the way to be industrially used in a wide range of irradiation conditions. (authors)

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

Preliminary concepts for detecting diversion of LWR spent fuel

International Nuclear Information System (INIS)

Sellers, T.A.

Sandia Laboratories, under the sponsorship of the Department of Energy, Office of Safeguards and Security, has been developing conceptual designs of advanced systems to rapidly detect diversion of LWR spent fuel. Three detection options have been identified and compared on the basis of timeliness of detection and cost. Option 1 is based upon inspectors visiting each facility on a periodic basis to obtain and review data acquired by surveillance instruments and to verify the inventory. Option 2 is based upon continuous inspector presence, aided by surveillance instruments. Option 3 is based upon the collection of data from surveillance instruments with periodic readout either at the facility or at a remote central monitoring and display module and occasional inspection. Surveillance instruments are included in each option to assure a sufficiently high probability of detection. An analysis technique with an example logic tree that was used to identify performance requirements is described. A conceptual design has been developed for Option 3 and the essential hardware elements are not being developed. These elements include radiation, crane and pool acoustic sensors, a Data Collection Module, a Local Collection Module, a Local Display Module and a Central Monitoring and Display Module. A demonstration, in operating facilities, of the overall system concept is planned for the March to June 1979 time frame

Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

2009-02-01

This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

Models for predicting fuel consumption in sagebrush-dominated ecosystems

Science.gov (United States)

Clinton S. Wright

2013-01-01

Fuel consumption predictions are necessary to accurately estimate or model fire effects, including pollutant emissions during wildland fires. Fuel and environmental measurements on a series of operational prescribed fires were used to develop empirical models for predicting fuel consumption in big sagebrush (Artemisia tridentate Nutt.) ecosystems....

FUEL CONSUMPTION EFFECT OF COMMERCIAL TURBOFANS ON GLOBAL WARMING

Energy Technology Data Exchange (ETDEWEB)

Onder Turan; T. Hikmet Karakoc [School of Civil Aviation, Anadolu University, Eskisehir (Turkey)

2008-09-30

The main objective pursued in this study is to parametrically investigate the fuel consumption effect of commercial turbofans on global warming. In this regard, Of the important parameters, specific fuel consumption of a commercial turbofans is taken into consideration. In order to minimize the effect of fuel consumption on global warming, the values of engine design parameters are optimized for maintaining minimum specific fuel consumption of high bypass turbofan engine under different flight conditions and design criteria. The backbones of optimization approach consisted of elitism-based genetic algorithm coupled with real parametric cycle analysis of a turbofan engine. For solving optimization problem a new software program is developed in MATLAB, while objective function is determined for minimizing the specific fuel consumption by considering the following parameters such as the fan pressure ratio ({pi}{sub f}), bypass ratio ({alpha}) and the fuel heating value [h{sub PR}-(kJ/kg)]. Accordingly, it may be concluded that the software program developed can successfully solve optimization problems at 1.2{le}{pi}{sub f}{le}2, 2{le}{alpha}{le}10 and 23000{le}h{sub PR}{le}120000 with aircraft flight Mach number {le}0.8. Fuel types used in preliminary engine cycle analysis were JP-4, JP-5, JP-8 and hydrogen in this paper.

Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

International Nuclear Information System (INIS)

Fairbairn, S.A.

1983-01-01

The present structure and intentions of the CEC sponsored cost sharing programme for LWR safety research are outlined. Detailed results are reported for two projects from this programme. The first project concerns experimental data on the thermohydraulic effects of flow diversion around ballooned fuel rods. Data are presented on single and two phase heat transfer in an electrically heated rod bundle. Detailed photographic data on droplet behaviour are also given. The second project is an investigation of the effects of zircaloy oxidation on rewetting during reflood. It is shown that as oxide thickness increases from 1μm to 76μm that rewet rates can increase by up to 40%. A systematic effect of oxidation on rewet temperatures is also noted. (author)

Extending dry storage of spent LWR fuel for up to 100 years

International Nuclear Information System (INIS)

Einziger, R.E.; McKinnon, M.A.; Machiels, A.J.

1999-01-01

Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and excessive

Extending dry storage of spent LWR fuel for up to 100 years

International Nuclear Information System (INIS)

Einziger, R. E.

1998-01-01

Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

ERDA LWR plant technology program: role of government/industry in improving LWR performance

International Nuclear Information System (INIS)

1975-01-01

Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program

Concept of innovative water reactor for flexible fuel cycle (FLWR)

International Nuclear Information System (INIS)

Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

2005-01-01

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

Wood fuels consumption in households in Montenegro

Directory of Open Access Journals (Sweden)

Glavonjić Branko D.

2013-01-01

Full Text Available The paper presents the results of researching wood fuels consumption in households in Montenegro in the heating season 2011/2012. The research was conducted in the period October-November 2012 on the sample of 5% of the total number of households which stated to use solid fuels for heating purposes in the 2011 census. Results of the conducted researches on the presence and amounts of fuels consumed in households in Montenegro showed that total firewood consumption in the heating season 2011/2012 (both urban and rural households was 703,571 m3. Wood consumption is the lowest in the households in the municipalities in the coastal zone, it is somewhat higher in central zone and the highest in the zone on the north of Montenegro. Average wood consumption in households in the coastal zone municipalities is 3.79 m3, in the central zone it is 5.02 m3 and on the north of Montenegro it is 6.74 m3/household. Observed on the level of Montenegro, average firewood consumption per household was 5.49 m3 and as such it best represents relatively low consumption level in the coastal zone and high consumption level on the north of Montenegro. Compared to the neighboring countries, average firewood consumption per household in Montenegro in the amount of 5.49 m3 is significantly lower than the average consumption in Serbia which is 7.3 m3/household as well as in Slovenia in the amount of 6.5 m3.

Investigation of Nuclide Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Broadhead, B.L.

1995-01-01

The radionuclide characteristics of light-water-reactor (LWR) spent fuel play key roles in the design and licensing activities for radioactive waste transportation systems, interim storage facilities, and the final repository site. Several areas of analysis require detailed information concerning the time-dependent behavior of radioactive nuclides including (1) neutron/gamma-ray sources for shielding studies, (2) fissile/absorber concentrations for criticality safety determinations, (3) residual decay heat predictions for thermal considerations, and (4) curie and/or radiological toxicity levels for materials assumed to be released into the ground/environment after long periods of time. The crucial nature of the radionuclide predictions over both short and long periods of time has resulted in an increased emphasis on thorough validation for radionuclide generation/depletion codes. Current radionuclide generation/depletion codes have the capability to follow the evolution of some 1600 isotopes during both irradiation and decay time periods. Of these, typically only 10 to 20 nuclides dominate contributions to each analysis area. Thus a quantitative ranking of nuclides over various time periods is desired for each of the analysis areas of shielding, criticality, heat transfer, and environmental dose (radiological toxicity). These rankings should allow for validation and data improvement efforts to be focused only on the most important nuclides. This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to

Fuel consumption from vehicles of China until 2030 in energy scenarios

International Nuclear Information System (INIS)

Zhang Qingyu; Tian Weili; Zheng Yingyue; Zhang Lili

2010-01-01

Estimation of fuel (gasoline and diesel) consumption for vehicles in China under different long-term energy policy scenarios is presented here. The fuel economy of different vehicle types is subject to variation of government regulations; hence the fuel consumption of passenger cars (PCs), light trucks (Lts), heavy trucks (Hts), buses and motor cycles (MCs) are calculated with respect to (i) the number of vehicles, (ii) distance traveled, and (iii) fuel economy. On the other hand, the consumption rate of alternative energy sources (i.e. ethanol, methanol, biomass-diesel and CNG) is not evaluated here. The number of vehicles is evaluated using the economic elastic coefficient method, relating to per capita gross domestic product (GDP) from 1997 to 2007. The Long-range Energy Alternatives Planning (LEAP) system software is employed to develop a simple model to project fuel consumption in China until 2030 under these scenarios. Three energy consumption decrease scenarios are designed to estimate the reduction of fuel consumption: (i) 'business as usual' (BAU); (ii) 'advanced fuel economy' (AFE); and (iii) 'alternative energy replacement' (AER). It is shown that fuel consumption is predicted to reach 992.28 Mtoe (million tons oil equivalent) with the BAU scenario by 2030. In the AFE and AER scenarios, fuel consumption is predicted to be 734.68 and 600.36 Mtoe, respectively, by 2030. In the AER scenario, fuel consumption in 2030 will be reduced by 391.92 (39.50%) and 134.29 (18.28%) Mtoe in comparison to the BAU and AFE scenarios, respectively. In conclusion, our models indicate that the energy conservation policies introduced by governmental institutions are potentially viable, as long as they are effectively implemented.

Wildland fire emissions, carbon, and climate: Modeling fuel consumption

Science.gov (United States)

Roger D. Ottmar

2014-01-01

Fuel consumption specifies the amount of vegetative biomass consumed during wildland fire. It is a two-stage process of pyrolysis and combustion that occurs simultaneously and at different rates depending on the characteristics and condition of the fuel, weather, topography, and in the case of prescribed fire, ignition rate and pattern. Fuel consumption is the basic...

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

International Nuclear Information System (INIS)

Loesoenen, Pekka

2002-01-01

A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

Plant for retention of 14C in reprocessing plants for LWR fuel elements

International Nuclear Information System (INIS)

Braun, H.; Gutowski, H.; Bonka, H.; Gruendler, D.

1983-01-01

The 14 C produced from nuclear power plants is actually totally emitted from nuclear power plants and reprocessing plants. Using the radiation protection principles proposed in ICRP 26, 14 C should be retained at heavy water moderated reactors and reprocessing plants due to a cost-benefit analysis. In the frame of a research work to cost-benefit analysis, which was sponsored by the Federal Minister of the Interior, an industrial plant for 14 C retention at reprocessing plants for LWR fuel elements has been planned according to the double alkali process. The double alkali process has been chosen because of the sufficient operation experience in the conventional chemical technique. In order to verify some operational parameters and to gain experiences, a cold test plant was constructed. The experiment results showed that the double alkali process is a technically suitable method with high operation security. Solidifying CaCO 3 with cement gives a product fit for final disposal

Psychological determinants of fuel consumption of purchased new cars

OpenAIRE

Peters, A; Gutscher, H; Scholz, R W

2011-01-01

With regard to reducing fuel consumption and CO2 emissions of road transport consumers' adoption of fuel-efficient vehicles is crucial. However, facing the ongoing trend of increasing car size and power, fuel consumption is apparently of lesser importance to most buyers. For the design of effective measures to change behavior and promote fuel-efficient cars, psychological factors should be considered. Drawing from psychological research on environmental behavior, we propose a model which inte...

Optimising Shovel-Truck Fuel Consumption using Stochastic ...

African Journals Online (AJOL)

Optimising the fuel consumption and truck waiting time can result in significant fuel savings. The paper demonstrates that stochastic simulation is an effective tool for optimising the utilisation of fossil-based fuels in mining and related industries. Keywords: Stochastic, Simulation Modelling, Mining, Optimisation, Shovel-Truck ...

LWR physics in SKODA Works

International Nuclear Information System (INIS)

Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.

1980-01-01

Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)

Alternatives for managing post LWR reactor nuclear wastes

International Nuclear Information System (INIS)

Platt, A.M.

1976-01-01

The two extremes in the LWR fuel cycle are discarding the spent fuel and recycling the U and Pu to the maximum extent possible. The waste volumes from the two alternatives are compared. A preliminary evaluation is made of the technology available for handling wastes from each step of the fuel cycle. The wastes considered are fuel materials, high--level wastes, other liquids, combustible and non-combustible solids, and non--high--level wastes. Evaluation of processing gaseous wastes indicates that technology is available for capture of Kr and I 2 , but further development is needed for T 2 . Technology for interim storage and geological isolation is considered adequate. An outline is given of the steps in the selection of a final storage site

Orbit selection of nanosatellite formation in term of fuel consumption

Science.gov (United States)

Pimnoo, Ammarin; Hiraki, Koju

In nanosatellite formation mission design, orbit selection is a necessary factor. Fuel consumption is also necessary to maintain the orbit. Therefore, the best orbit should be the one of minimum fuel consumption for nanosatellite formation. The purpose of this paper is to provide a convenient way to estimate fuel consumption for a nanosatellite to keep formation flying. The formation is disturbed by J _{2} perturbation and other perturbing accelerations. Firstly, the Hill-Clohessy-Wiltshire equations are used in the analysis. Gaussian variation of parameters is included into the Hill’s equation to analyze the variation of Kaplerian orbital elements. The J _{2} perturbation and other perturbing accelerations such as atmospheric drag, solar-radiation pressure and third-body perturbations are considered. Thus, a linear model based on Hill’s equation is established to estimate fuel consumption. Finally, an example of the best orbit for formation flying with minimum fuel consumption shall be presented.

Modelling of a LWR open fuel cycle using the message

Energy Technology Data Exchange (ETDEWEB)

Estanislau, Fidéllis B.G.L. e; Jonusan, Raoni A.S.; Costa, Antonella L.; Pereira, Claubia, E-mail: fidellis01@hotmail.com, E-mail: rjonusan@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

The main goal of the national energy planning is the development of a short and long-term strategies based on a holistic evaluation of all available energy sources guiding trends and delimiting expansion alternatives in the energetic sector. For a better understanding of the future possibilities, energy systems analyses are indispensable and support in the decision making related to the long term strategy and energy planning. Due to the projections for increased energy consumption according to the Energy Decennial Plan (year 2015) and the need to reduce greenhouse gas emissions presented by Brazil in the UNFCCC (United Nations Framework Convention on Climate Change), alternative energy sources such as solar, wind, nuclear and biomass sources have played an important role in the world energy matrix. In this way, since the nuclear energy is an option for the national energy mix, the present work aims to use the modelling tool MESSAGE (Model for Energy Supply System Alternatives and Their General Environmental Impact) to analyze and evaluate a nuclear power plant in an energy system. This tool is an optimization model for medium and long-term energy planning taking into account conversion and distribution technologies, energy policies and scenarios to satisfy a determined demand and systems constraints. In this work, a reproduction of results considering an LWR (Light Water Reactor) open-cycle are presented using a model in the MESSAGE code. (author)

Nondestructive analysis of irradiated fuels

International Nuclear Information System (INIS)

Dudey, N.D.; Frick, D.C.

1977-01-01

The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

Cost Savings of Nuclear Power with Total Fuel Reprocessing

International Nuclear Information System (INIS)

Solbrig, Charles W.; Benedict, Robert W.

2006-01-01

The cost of fast reactor (FR) generated electricity with pyro-processing is estimated in this article. It compares favorably with other forms of energy and is shown to be less than that produced by light water reactors (LWR's). FR's use all the energy in natural uranium whereas LWR's utilize only 0.7% of it. Because of high radioactivity, pyro-processing is not open to weapon material diversion. This technology is ready now. Nuclear power has the same advantage as coal power in that it is not dependent upon a scarce foreign fuel and has the significant additional advantage of not contributing to global warming or air pollution. A jump start on new nuclear plants could rapidly allow electric furnaces to replace home heating oil furnaces and utilize high capacity batteries for hybrid automobiles: both would reduce US reliance on oil. If these were fast reactors fueled by reprocessed fuel, the spent fuel storage problem could also be solved. Costs are derived from assumptions on the LWR's and FR's five cost components: 1) Capital costs: LWR plants cost $106/MWe. FR's cost 25% more. Forty year amortization is used. 2) The annual O and M costs for both plants are 9% of the Capital Costs. 3) LWR fuel costs about 0.0035 $/kWh. Producing FR fuel from spent fuel by pyro-processing must be done in highly shielded hot cells which is costly. However, the five foot thick concrete walls have the advantage of prohibiting diversion. LWR spent fuel must be used as feedstock for the FR initial core load and first two reloads so this FR fuel costs more than LWR fuel. FR fuel costs much less for subsequent core reloads ( 6 /MWe. The annual cost for a 40 year licensed plant would be 2.5 % of this or less if interest is taken into account. All plants will eventually have to replace those components which become radiation damaged. FR's should be designed to replace parts rather than decommission. The LWR costs are estimated to be 2.65 cents/kWh. FR costs are 2.99 cents/kWh for the first

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Program plan for research and development in support of LWR fuel recycle

International Nuclear Information System (INIS)

1975-01-01

The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

Perspectives on fuel ethanol consumption and trade

International Nuclear Information System (INIS)

Walter, Arnaldo; Dolzan, Paulo; Piacente, Erik; Borges da Cunha, Kamyla; Rosillo-Calle, Frank

2008-01-01

Since the year 2000 or so there has been a rapid growth on fuel ethanol production and consumption, particularly in US and Brazil. Ethanol trade represented about 10% of world consumption in 2005, Brazil being the main exporter. The most important consumer markets - US and European Union (EU) - have trade regimes that constrained the comparative advantages of the most efficient producers, such as Brazil. This paper evaluates the fuel ethanol market up to 2030 together with the potential for international biotrade. Based on forecasts of gasoline consumption and on targets and mandates of fuel ethanol use, it is estimated that demand could reach 272 Gl in 2030, displacing 10% of the estimated demand of gasoline (Scenario 1), or even 566 Gl in the same year, displacing about 20% of the gasoline demand (Scenario 2). The analysis considers fuel ethanol consumption and production in US, EU-25, Japan, China, Brazil and the rest of the world (ROW-BR). Without significant production of ethanol from cellulosic materials in this period, displacing 10% of the gasoline demand in 2030, at reasonable cost, can only be accomplished by fostering fuel ethanol production in developing countries and enhancing ethanol trade. If the US and EU-25 reach their full production potential (based on conventional routes), the minimum amount that could be traded in 2030 would be about 34 Gl. Displacing 20% of the gasoline demand by 2030 will require the combined development of second-generation technologies and large-scale international trade in ethanol fuel. Without second-generation technologies, Scenario 2 could become a reality only with large-scale production of ethanol from sugarcane in developing countries, e.g., Brazil and ROW-BR could be able to export at least 14.5 Gl in 2010, 73.9 Gl in 2020 and 71.8 Gl in 2030. (author)

In-core materials testing under LWR conditions in the Halden reactor

International Nuclear Information System (INIS)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

2002-01-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

In-core materials testing under LWR conditions in the Halden reactor

Energy Technology Data Exchange (ETDEWEB)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

2002-07-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

International Nuclear Information System (INIS)

Tulenko, J.S.; Connell, R.G.

1995-07-01

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. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the 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 resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program 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 Metal Bonded LWR fuel

Swiss R and D on uranium-free LWR fuels for plutonium incineration

International Nuclear Information System (INIS)

Stanculescu, A.; Chawla, R.; Degueldre, C.; Kasemeyer, U.; Ledergerber, G.; Paratte, J.M.

1999-01-01

The most efficient way to enhance the plutonium consumption in LWRs is to eliminate plutonium production altogether. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. The inert matrix material studied at PSI is zirconium oxide. For reactivity control reasons, adding a burnable poison to this fuel proves to be necessary. The studies performed at PSI have identified erbium oxide as the most suitable candidate for this purpose. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of the chosen single phase zirconium/ erbium/plutonium oxide material stabilised as a cubic solution by yttrium. The results to-date, obtained for inert matrix samples containing thorium or cerium as plutonium substitute, confirm the robustness and stability of this material. With regard to reactor physics aspects, our studies indicate the feasibility of uranium-free, plutonium-fuelled cores having operational characteristics quite similar to those of conventional UO 2 -fuelled ones, and much higher plutonium consumption rates, as compared to 100% MOX loadings. The safety features of such cores, based on results obtained from static neutronics calculations, show no cliff edges. However, the need for further detailed transient analyses is clearly recognised. Summarising, PSI's studies indicate the feasibility of a uranium-free plutonium fuel to be considered in 'maximum plutonium consumption LWRs' operating in a 'once-through' mode. With regard to reactor physics, future efforts will concentrate on strengthening the safety case of uranium-free cores, as well as on improving the integral data base for validation of the neutronics calculations. Material technology studies will be continued to investigate the physico-chemical properties of the inert matrix fuel containing plutonium and will focus on the planning and evaluation of

Spent fuel characteristics provided by the CDB: An update

International Nuclear Information System (INIS)

Notz, K.J.; Salmon, R.; Welch, T.D.; Reich, W.J.; Moore, R.S.

1992-01-01

The Characteristics Data Base (CDB) task provides OCRWM with the detailed technical characteristics of potential repository wastes, which consist primarily of commercial spent nuclear fuel, but also includes other spent fuel (and also high-level and miscellaneous wastes). A major revision of the original CDB report and PC data bases has just been completed under formal QA peer review guidelines and Revision 1 is ready to be issued. This paper describes the classification scheme developed for LWR fuel assemblies and the five PC data bases for LWR spent fuel, which provide data on quantities, assemblies, radiological properties, non-fuel assembly hardware, and serial numbers. The future role of other (i.e., non-LWR) spent fuel is also cited

Summary of Research on Light Water Reactor Improvement Concepts

International Nuclear Information System (INIS)

Mowery, Alfred L.

2002-01-01

The Arms Control and Disarmament Agency of the U.S. Department of State instituted a study aimed at improving the light water reactor (LWR) fuel consumption efficiency as an alternative to fuel recycle in the late 1970s. Comparison of the neutron balance tables of an LWR (1982 design) and an 'advanced' Canada deuterium uranium (CANDU) reactor explained that the relatively low fuel efficiency of the LWR was not primarily a consequence of water moderator absorptions. Rather, the comparatively low LWR fuel efficiency resulted from its use of poison to hold down startup reactivity together with other neutron losses. The research showed that each neutron saved could reduce fuel consumption by about 5%. In a typical LWR some 5 neutrons (out of 100) were absorbed in control poisons over a cycle. There are even more parasitic and leakage neutron absorptions. The objective of the research was to find ways to minimize control, parasitic, and other neutron losses aimed at improved LWR fuel consumption. Further research developed the concept of 'putting neutrons in the bank' in 238 U early in life and 'drawing them out of the bank' late in life by burning the 239 Pu produced. Conceptual designs were explored that could both control the reactor and substantially improve fuel efficiency and minimize separative work requirements.The U.S. Department of Energy augmented its high burnup fuel program based on the research in the late 1970s. As a result of the success of this program, fuel burnup in U.S. LWRs has almost doubled in the intervening two decades

77 FR 18718 - Petroleum Reduction and Alternative Fuel Consumption Requirements for Federal Fleets

Science.gov (United States)

2012-03-28

... Petroleum Reduction and Alternative Fuel Consumption Requirements for Federal Fleets AGENCY: Office of...-required reduction in petroleum consumption and increase in alternative fuel consumption for Federal fleets... regulations for a statutorily-required reduction in petroleum consumption and increase in alternative fuel...

Real driving emissions and fuel consumption characteristics of Istanbul public transportation

Directory of Open Access Journals (Sweden)

Ozener Orkun

2017-01-01

Full Text Available Public transportation, which uses intra city lines frequently, has vital importance on the cities air pollution. The fossil fuel based drive units, which emits pollutants, are the primary source of this interest. Also, the fuel consumption is another major concern because of economic aspects. For an efficient and clear transportation, the pollutants and fuel consumption has to be analyzed, considering the operating conditions. In this context, the Metrobus line of Istanbul city which crosses from European side to Asian side of the city was analyzed with portable emission measurement system and portable fuel consumption meter devices. The relevant bus operating data were also collected during the operation. The data were analyzed while considering the operating modes like acceleration, deceleration, and constant speed cruises. The emission factors were developed. The pollutant emissions generally decreased as the vehicle speed increased while the fuel consumption increased for the same acceleration level. These results show the importance of operating conditions and their non-linear effect on emissions and fuel consumption Istanbul public transportation.

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

International Nuclear Information System (INIS)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document

Experimental Assessment of a New Passive Neutron Multiplication Counter for Partial Defect Verification of LWR Fuel Assemblies

International Nuclear Information System (INIS)

LaFleur, A.; Menlove, H.; Park, S.-H.; Lee, S. K.; Oh, J.-M.; Kim, H.-D.

2015-01-01

The development of non-destructive assay (NDA) capabilities to improve partial defect verification of spent fuel assemblies is needed to improve the timely detection of the diversion of significant quantities of fissile material. This NDA capability is important to the implementation of integrated safeguards for spent fuel verification by the International Atomic Energy Agency (IAEA) and would improve deterrence of possible diversions by increasing the risk of early detection. A new NDA technique called Passive Neutron Multiplication Counter (PNMC) is currently being developed at Los Alamos National Laboratory (LANL) to improve safeguards measurements of LightWater Reactor (LWR) fuel assemblies. The PNMC uses the ratio of the fast-neutron emission rate to the thermalneutron emission rate to quantify the neutron multiplication of the item. The fast neutrons versus thermal neutrons are measured using fission chambers (FC) that have differential shielding to isolate fast and thermal energies. The fast-neutron emission rate is directly proportional to the neutron multiplication in the spent fuel assembly; whereas, the thermalneutron leakage is suppressed by the fissile material absorption in the assembly. These FCs are already implemented in the basic Self-Interrogation Neutron Resonance Densitometry (SINRD) detector package. Experimental measurements of fresh and spent PWR fuel assemblies were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using a hybrid PNMC and SINRD detector. The results from these measurements provides valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. (author)

Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

International Nuclear Information System (INIS)

Boer, B.; Sen, R.S.; Pope, M.A.; Ougouag, A.M.

2011-01-01

The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO 2 kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 (micro)mm kernel diameter, 100 (micro)mm buffer, 35 (micro)mm IPyC, 35 (micro)mm SiC, 40 (micro)mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10 -2 failure probability. For a 'best-estimate' FGR fraction of 50

Power generation versus fuel production in light water hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1977-06-01

The economic potentials of fissile-fuel-producing light-water hybrid reactors (FFP-LWHR) and of fuel-self-sufficient (FSS) LWHR's are compared. A simple economic model is constructed that gives the capital investment allowed for the hybrid reactor so that the cost of electricity generated in the hybrid based energy system equals the cost of electricity generated in LWR's. The power systems considered are LWR, FSS-LWHR, and FFP-LWHR plus LWR, both with and without plutonium recycling. The economic potential of FFP-LWHR's is found superior to that of FSS-LWHR's. Moreover, LWHR's may compete, economically, with LWR's. Criteria for determining the more economical approach to hybrid fuel or power production are derived for blankets having a linear dependence between F and M. The examples considered favor the power generation rather than fuel production

A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

International Nuclear Information System (INIS)

Blomeke, J.O.

1988-06-01

The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

Fuel consumption in the transport of technical broadleaf roundwood in lowland areas

Directory of Open Access Journals (Sweden)

Danilović Milorad

2015-01-01

Full Text Available This paper presents the results of an analysis of fuel consumption in the transport of technical roundwood of soft broadleaves from the felling site to a roadside landing using forwarders and tractor assemblies. The research was performed in various operating conditions in the area of FE "Banat" Pančevo. On the basis of the results of the analysis of variance, the data recorded in a variety of conditions were grouped. In addition, the dependence of fuel consumption on the average volume of tour was estimated. The results of the conducted analysis indicate that operating conditions significantly affect fuel consumption of the investigated vehicles. The elements of statistical analysis of the dependence of fuel consumption on the volume of load indicate that an increase in load causes increased fuel consumption per unit of production. Having in mind the results of the analysis of variance, unique norms of fuel consumption were adopted for practical purposes. The highest average consumption (1.21 L/m3 was achieved by a tractor assembly (Same Laser 130 tractor and Imako TP12 trailer with a Loglift 61F hydraulic crane, while significantly lower consumption was achieved by a John Deere 1210E forwarder (1.06 L/m3. In favourable operating conditions, consumption of the forwarder was about 0.9 L/m3.

Core design of super LWR with double tube water rods

International Nuclear Information System (INIS)

Wu, Jianhui; Oka, Yoshiaki

2014-01-01

Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

In-use vs. type-approval fuel consumption of current passenger cars in Europe

International Nuclear Information System (INIS)

Ntziachristos, L.; Mellios, G.; Tsokolis, D.; Keller, M.; Hausberger, S.; Ligterink, N.E.; Dilara, P.

2014-01-01

In-use fuel consumption data of 924 passenger cars (611 petrol, 313 diesel) were collected from various European sources and were evaluated in comparison to their corresponding type-approval values. The analysis indicated that the average in-use fuel consumption was higher than the type-approval one by 11% for petrol cars and 16% for diesel cars. Comparison of this dataset with the Travelcard database in the Netherlands showed that the deviation increased for late model years and in particular for cars with low type-approval values. The deviation was higher than 60% for vehicles registered in 2012 within the 90–100 gCO 2 /km bin. Unrealistic vehicle resistances used in type-approval were identified as one of the prime reasons of the difference. A simplified linear model developed in the study may be used to predict in-use fuel consumption based on data publicly available. The model utilizes the fuel consumption measured in type-approval, the mass, and the engine capacity to provide in-use fuel consumption. This may be either used to correct fuel consumption factors currently utilized by emission models (e.g. COPERT, HBEFA, VERSIT+, and others) or could be used independently to make projections on how fuel consumption may develop on the basis of changing future passenger cars characteristics. - Highlights: • In-use fuel consumption of petrol and diesel passenger cars is 11% and 16% higher than type-approval, respectively. • The relative difference between in-use and type-approval increases for late model and vehicles with low consumption. • Unrealistically low vehicle resistances are identified as a prime reason of low type-approval fuel consumption. • A model developed predicts in-use consumption on the basis of type-approval consumption, vehicle mass, and engine capacity

Prediction of fuel consumption of mining dump trucks: A neural networks approach

International Nuclear Information System (INIS)

Siami-Irdemoosa, Elnaz; Dindarloo, Saeid R.

2015-01-01

Highlights: • A neural network model of fuel consumption in mining haul trucks was constructed and tested. • Using the cyclic activities, the model was able to predict unseen (testing) data. • Trucks idle times were identified as the most important unnecessary energy consuming portion of the network. • Practical remedies, based on the nature of mining operations, were proposed to reduce the energy consumption. - Abstract: Fuel consumption of mining dump trucks accounts for about 30% of total energy use in surface mines. Moreover, a fleet of large dump trucks is the main source of greenhouse gas (GHG) generation. Modeling and prediction of fuel consumption per cycle is a valuable tool in assessing both energy costs and the resulting GHG generation. However, only a few studies have been published on fuel prediction in mining operations. In this paper, fuel consumption per cycle of operation was predicted using artificial neural networks (ANN) technique. Explanatory variables were: pay load, loading time, idled while loaded, loaded travel time, empty travel time, and idled while empty. The output variable was the amount of fuel consumed in one cycle. Mean absolute percentage error (MAPE) of 10% demonstrated applicability of ANN in prediction of the fuel consumption. The results demonstrated the considerable effect of mining trucks idle times in fuel consumption. A large portion of the unnecessary energy consumption and GHG generation, in this study, was solely due to avoidable idle times. This necessitates implementation of proper actions/remedies in form of both preventive and corrective actions

Domestic fuel energy consumption in an Indian urban ecosystem

International Nuclear Information System (INIS)

Misra, M.K.; Rao, B.G.; Nisanka, S.K.; Sahu, N.C.

1995-01-01

Berhampur is one of the oldest and biggest towns of Orissa State where 93% of people earn their living from non-primary economic activities. The fuel consumption structure reveals an element of rural bias since the domestic sector derives 49% of its energy used for cooking and heating from biomass sources. Firewood is the only fuel used by all income groups and almost all occupations. Other traditional fuels such as dungcake, agricultural waste and leaf litter are used in the town with different degrees of dominance in various occupation groups and income classes. The inequality of income distribution is reflected in the fuel consumption structure of the rich and poor with greater dependence on non-biomass in the high income group and biomass in the low income group. Family size is an important determinant of biomass fuel use. For the urban poor, there is a relationship between income and fuel consumption. There exists scope for enlarging the biomass base by utilising the available bio-wastes and developing a green belt around the town. This requires some rural features in the urban area using new technologies. (author)

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

Energy Technology Data Exchange (ETDEWEB)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

Cross-section covariance propagation for LWR fuel cells in one and two dimensions - 308

International Nuclear Information System (INIS)

Ball, M.; Novog, D.R.; Parisi, C.; D'Auria, F.

2010-01-01

Within the framework of the Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of LWRs Benchmark sponsored by the OECD/NEA, a tool has been developed for the propagation of covariance uncertainty through resonance self-shielding and other neutron kinetics calculations using a direct, cross-section generation and substitution approach. The motivation behind the work described in this paper was to develop a portable uncertainty propagation tool that could be easily implemented with several neutron kinetics codes, without relying on detailed knowledge of the internal workings of those codes or access to adjoint solutions. Implemented initially with the SCALE code package, 'self-shielded' covariance matrices for common LWR fuel cells have been calculated, as well as contributions to K eff uncertainty by selected neutron cross-sections and processes in both one and two dimensions. The one dimensional results generated by the tool are compared against those obtained using the TSUNAMI-1D module of SCALE in order to verify the efficacy of the methodology. One-dimensional results show good agreement with TSUNAMI-1D, but there is also an indication that the loss of dimensionality corresponding to one-dimensional equivalent geometries of two-dimensional fuel cells may lead to significant changes in the calculated uncertainty on K eff arising from particular neutron-nuclide reactions. (authors)

Characterisation of high-burnup LWR fuel rods through gamma tomography

International Nuclear Information System (INIS)

Caruso, S.

2007-01-01

Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235 U enrichment of the fresh UO 2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO 2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

Materials behavior in interim storage of spent fuel

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

1982-01-01

Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

International Nuclear Information System (INIS)

Linde, A. van der.

1989-04-01

Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

Light Water Reactor (LWR) safety

International Nuclear Information System (INIS)

Sehgal, Bal Raj

2006-01-01

In this paper, a historical review of the developments in the safely of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3 + LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA. however, they are valid for the rest of the Western World. This review can not do justice to the many many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above

Fuel consumption and greenhouse gas calculator for diesel and biodiesel-powered vehicles

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

Factors that influence fuel consumption include environmental conditions, maintenance, poor driving techniques, and driving speed. Developed by Natural Resources Canada, the SmartDriver training programs were designed to help fleet managers, drivers, and instructors to learn methods of improving fuel economy. This fuel consumption and greenhouse gas (GHG) calculator for diesel and biodiesel-powered vehicles provides drivers with a method of calculating fuel consumption rates when driving. It includes a log-book in which to record odometer readings and a slide-rule in which to determine the litres of fuel used during a trip. The scale showed the number of kg of GHGs produced by burning a particular amount of fuel for both biodiesel and diesel fuels. 1 fig.

MOX fuel reprocessing and recycling

International Nuclear Information System (INIS)

Guillet, J.L.

1990-01-01

This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

Fuel consumption of gasoline ethanol blends at different engine rotational

Directory of Open Access Journals (Sweden)

Y. Barakat

2016-09-01

Full Text Available Fuel consumption (mf kg/h was estimated for two hydrocarbon gasolines (BG1-OE and BG2-OE and their ethanol blends which contain from 4 to 20 vol.% of ethanol. Fuel consumption experiments for sixteen fuel samples (5 L each, were conducted on a four cylinder, four stroke spark ignition test vehicle Sahin car, Type 1.45, model 2001. The engine has a swept volume of 1400 c.c., a compression ratio of 8.3:1 and a maximum power of 78 HP at 5500 rpm. The obtained data reveal that the relation between fuel consumption and ethanol concentration is linear. Six linear equations for BG1-ethanol blends and BG2-ethanol ones at the investigated rotational speeds, were developed. Fuel consumption values of the first set of gasoline-ethanol blends are lower than that of the second set. This may be attributed to the difference in the chemical composition of base gasolines BG1 in the first set which is enriched in the less volatile reformate if compared with the second set which is more enriched in isomerate, the more volatile refinery stream.

Wastes from selected activities in two light-water reactor fuel cycles

International Nuclear Information System (INIS)

Palmer, C.R.; Hill, O.F.

1980-07-01

This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume

Low-decontamination approach to a proliferation-resistant fuel cycle

International Nuclear Information System (INIS)

Asquith, J.G.; Grantham, L.F.

1978-01-01

To prevent the diversion of nuclear material from power production to weapon production either by a nation or by clandestine groups within a nation, the nuclear fuel cycle must be proliferation-resistant and safeguarded. Potentially proliferation-resistant and safeguarded fuel cycles based on low-decontamination pyroreprocessing have been developed for the light water reactor (LWR), fast breeder reactor (FBR), and FBR-LWR combination. The major penalty for recycling fission products to the LWR is that fuel enrichment must be somewhat greater to overcome parasitic fission product absorption of neutrons. In the FBR, the major penalty is a slight reduction in breeding ratio due to the displacement of fertile material by fission products. Preliminary cost analysis indicates that these fuel cycles are economically competitive with fuel cycles using conventional reprocessing or those using virgin uranium if spent fuel storage costs are considered

Analysis of Virtual Sensors for Predicting Aircraft Fuel Consumption

Data.gov (United States)

National Aeronautics and Space Administration — Previous research described the use of machine learning algorithms to predict aircraft fuel consumption. This technique, known as Virtual Sensors, models fuel...

Experience gained in the current LWR that influence the design and operation of the LWR advanced from the viewpoint of safety analysis

International Nuclear Information System (INIS)

Barrera, J.; Corisco, M.; Riverola, J.

2010-01-01

Since the construction of the first light water reactors (LWR) safety analysis has played a very important role in the operation and its evolution to come up with designs that are currently operating. With new tools available, this role will see increased allowing more efficient operation with security assessments in real time, and a more efficient designs both in terms of fuel efficiency and from the security of the plant during operation.

Accessing and constructing driving data to develop fuel consumption forecast model

Science.gov (United States)

Yamashita, Rei-Jo; Yao, Hsiu-Hsen; Hung, Shih-Wei; Hackman, Acquah

2018-02-01

In this study, we develop a forecasting models, to estimate fuel consumption based on the driving behavior, in which vehicles and routes are known. First, the driving data are collected via telematics and OBDII. Then, the driving fuel consumption formula is used to calculate the estimate fuel consumption, and driving behavior indicators are generated for analysis. Based on statistical analysis method, the driving fuel consumption forecasting model is constructed. Some field experiment results were done in this study to generate hundreds of driving behavior indicators. Based on data mining approach, the Pearson coefficient correlation analysis is used to filter highly fuel consumption related DBIs. Only highly correlated DBI will be used in the model. These DBIs are divided into four classes: speed class, acceleration class, Left/Right/U-turn class and the other category. We then use K-means cluster analysis to group to the driver class and the route class. Finally, more than 12 aggregate models are generated by those highly correlated DBIs, using the neural network model and regression analysis. Based on Mean Absolute Percentage Error (MAPE) to evaluate from the developed AMs. The best MAPE values among these AM is below 5%.

REDUCTION OF FUEL CONSUMPTION BY USINGAERODYNAMIC DOOR TOP PROFILE ON TRAILERAND SEMI-TRAILER

OpenAIRE

Topçu, İlhan; Çolakoğlu, Gülsüm Yeşim; Keleş, Uğur

2016-01-01

Fuel efficiency in vehicles is the important issue that dwells on for recent years not only for consumed fuel cost but also for the reduction of pollution caused by emission. There are several factor which affect the fuel consumption. One of them is aerodynamic resistance. Any decrease in the aerodynamic resistance has a great importance at fuel consumption. Especially fuel consumption in tractor trailer combination is much more than other automobiles. The main purpose of this report is to ex...

Heavy truck modeling for fuel consumption. Simulations and measurements

Energy Technology Data Exchange (ETDEWEB)

Sandberg, T.

2001-12-01

Fuel consumption for heavy trucks depends on many factors like roads, weather, and driver behavior that are hard for a manufacturer to influence. However, one design possibility is the power train configuration. Here a new simulation program for heavy trucks is created to find the configuration of the power train that gives the lowest fuel consumption for each transport task. For efficient simulations the model uses production code for speed and gear control, and it uses exchangeable data sets to allow simulation of the whole production range of engine types, on recorded road profiles from all over the world. Combined with a graphical user interface this application is called STARS (Scania Truck And Road Simulation). The forces of rolling resistance and air resistance in the model are validated through an experiment where the propeller shaft torque of a heavy truck is measured. It is found that the coefficient of rolling resistance is strongly dependent on tire temperature, not only on vehicle speed as expected. This led to the development of a new model for rolling resistance. The model includes the dynamic behavior of the tires and relates rolling resistance to tire temperature and vehicle speed. In another experiment the fuel consumption of a test truck in highway driving is measured. The altitude of the road is recorded with a barometer and used in the corresponding simulations. Despite of the limited accuracy of this equipment the simulation program manage to predict a level of fuel consumption only 2% lower than the real measurements. It is concluded that STARS is a good tool for predicting fuel consumption for trucks in highway driving and for comparing different power train configurations.

Fuel consumption rates of passenger cars in China: Labels versus real-world

International Nuclear Information System (INIS)

Huo Hong; Yao Zhiliang; He Kebin; Yu Xin

2011-01-01

Recently, China has implemented many policy measures to control the oil demand of on-road vehicles. In 2010, China started to report the fuel consumption rates of light-duty vehicles tested in laboratory and to require new vehicles to show the rates on window labels. In this study, we examined the differences between the test and real-world fuel consumption of Chinese passenger cars by using the data reported by real-world drivers on the internet voluntarily. The sales-weighted average fuel consumption of new cars in China in 2009 was 7.80 L/100 km in laboratory and 9.02 L/100 km in real-world, representing a difference of 15.5%. For the 153 individual car models examined, the real-world fuel consumption rates were -8 to 60% different from the test values. The simulation results of the International Vehicle Emission model show that the real-world driving cycles in 22 selected Chinese cities could result in -8 to 34% of changes in fuel consumption compared to the laboratory driving cycle. Further government effort on fuel consumption estimates adjustment, local driving cycle development, and real-world data accumulation through communication with the public is needed to improve the accuracy of the labeling policy. - Research highlights: → China requires new cars sold to show the test fuel use levels on window labels. → Real-world fuel consumption rates of cars are 15.5% higher than the label values. → Discrepancy between the test cycle and real conditions is a major cause for the gap. → China should adjust the estimates, collect fuel use data, and develop driving cycles. → More official and academic efforts are needed to improve the labeling policy.

LWR high burn-up operation and MOX introduction. Fuel cycle performance from the viewpoint of waste management

International Nuclear Information System (INIS)

Inagaki, Yaohiro; Iwasaki, Tomohiko; Niibori, Yuichi; Sato, Seichi; Ohe, Toshiaki; Kato, Kazuyuki; Torikai, Seishi; Nagasaki, Shinya; Kitayama, Kazumi

2009-01-01

From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO 2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO 3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO 2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO 2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO 3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO 3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance. (author)

Methods of rational fuel consumption by the dump trucks in deep pits

Directory of Open Access Journals (Sweden)

Yu. I. Lel’

2017-12-01

Full Text Available The detailed analysis of “consumption Norms of fuel and lubricants...” has been performed. It was approved by the Ministry of transport of the Russian Federation and it is an official document for the valuation and planning of fuel consumption in road transport, including open pits. It was found that the “rate...” was deficient in the specific operating conditions of mining dump trucks and a significant discrepancy with the actual fuel consumption. To the greatest extent, these shortcomings are noticeable in deep pits. In the result of experimental and analytical studies a new method of rationing the diesel fuel consumption of mining dump trucks was developed. It is based on the calculation of fuel consumption for vehicle cycle and bringing the actual path length to the conventional horizontal transportation distances using the criterion of energy. This method requires comprehensive consideration of the design parameters of mining dump trucks – mining and road traffic conditions. As a criterion for evaluating the difficulty of transporting it justifies the use of the coefficient of difficulty of the course, showing how many times the energy consumption in the movement of trucks on the highway exceed the real energy consumption while driving on a horizontal road with the crushed-stone surface of the same length. The formulae are derived to determine the given distance of transportation on the basis of the horizontal equivalent of the vertical movement of the rock mass. It was found that the linear rate of flow of diesel fuel to mileage and transportation work under other equal conditions is determined by two main factors: the distance of transportation of L and the coefficient μ of difficulty of the course. As an example, the calculation of the linear differentiated norms of fuel consumption of dump trucks CAT 785B in an “Udachny” pit AK “ALROSA” (JSC. The developed method provides high accuracy of regulation of diesel fuel

Consumption of fuels and energy in the coking industry and ways of reducing consumption

Energy Technology Data Exchange (ETDEWEB)

Vasil' ev, Yu.S.; Tsel' ik, M.P.; Belkina, T.V. (Khar' kovskii Nauchno-Issledovatel' skii Uglekhimicheski Institut (USSR))

1989-08-01

Coking plants in the USSR consume 4,000 million kWh electric energy, 100 million GJ heat energy and 35,000 million m{sup 3} gaseous fuels per year. Structure of energy consumption is the following: 68% gaseous fuels, 24% steam and 8% electric energy. Processes of coal preparation, crushing, mixing, coking and quenching are analyzed from the point of view of energy consumption. The following methods for reducing energy consumption are discussed: using the FM-25 flotation machines for flotation of coking coal slurries, briquetting the whole coal charge for coking, optimization of air supply rates for combustion of gases used for coke oven heating, use of control systems for coke oven heating considering coal charge density, waste heat utilization from quenching. 4 refs.

Thermal conductivity of heterogeneous LWR MOX fuels

Science.gov (United States)

Staicu, D.; Barker, M.

2013-11-01

It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

Ramp metering with an objective to reduce fuel consumption

OpenAIRE

Vreeswijk, Jacob Dirk; Woldeab, Zeremariam; de Koning, Anne; Bie, Jing

2011-01-01

Ramp meters successfully decrease congestion but leave a burden on the traffic situation at on-ramps. Chaotic queuing leads to many stop-and-go movements and causes inefficiency where fuel consumption is concerned. As part of the eCoMove project, complementary strategies are being designed and evaluated to reduce fuel consumption at metered on-ramps, using vehicle-to-infrastructure communication. This paper presents the design of two strategies, as well as their effect as derived from simulat...

Arrival condition of spent fuel after storage, handling, and transportation

International Nuclear Information System (INIS)

Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

1982-11-01

This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Abe, Tomoyuki; Kashimura, Takuo; Nagai, Toshihisa; Maeda, Seichiro; Yamaguchi, Toshiya; Kuroki, Ryoichiro

2003-07-01

The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

Impact of methanol-gasoline fuel blend on the fuel consumption and exhaust emission of a SI engine

Science.gov (United States)

Rifal, Mohamad; Sinaga, Nazaruddin

2016-04-01

In this study, the effect of methanol-gasoline fuel blend (M15, M30 and M50) on the fuel consumption and exhaust emission of a spark ignition engine (SI) were investigated. In the experiment, an engine four-cylinder, four stroke injection system (engine of Toyota Kijang Innova 1TR-FE) was used. Test were did to know the relation of fuel consumption and exhaust emission (CO, CO2, HC) were analyzed under the idle throttle operating condition and variable engine speed ranging from 1000 to 4000 rpm. The experimental result showed that the fuel consumption decrease with the use of methanol. It was also shown that the CO and HC emission were reduced with the increase methanol content while CO2 were increased.

Monitoring of bunker fuel consumption

Energy Technology Data Exchange (ETDEWEB)

Faber, J.; Nelissen, D.; Smit, M.

2013-03-15

Monitoring of fuel consumption and greenhouse gas emissions from international shipping is currently under discussion at the EU level as well as at the IMO (International Maritime Organization). There are several approaches to monitoring, each with different characteristics. Based on a survey of the literature and information from equipment suppliers, this report analyses the four main methods for monitoring emissions: (1) Bunker delivery notes (i.e. a note provided by the bunker fuel supplier specifying, inter alia, the amount of fuel bunkered); (2) Tank sounding (i.e. systems for measuring the amount of fuel in the fuel tanks); (3) Fuel flow meters (i.e. systems for measuring the amount of fuel supplied to the engines, generators or boilers); and (4) Direct emissions monitoring (i.e. measuring the exhaust emissions in the stack). The report finds that bunker delivery notes and tank soundings have the lowest investment cost. However, unless tank sounding is automated, these systems have higher operational costs than fuel flow meters or direct emissions monitoring because manual readings have to be entered in monitoring systems. Fuel flow meters have the highest potential accuracy. Depending on the technology selected, their accuracy can be an order of magnitude better than the other systems, which typically have errors of a few percent. By providing real-time feed-back on fuel use or emissions, fuel flow meters and direct emissions monitoring provide ship operators with the means to train their crew to adopt fuel-efficient sailing methods and to optimise their maintenance and hull cleaning schedules. Except for bunker delivery notes, all systems allow for both time-based and route-based (or otherwise geographically delineated) systems.

Nonresidential buildings energy consumption survey: 1979 consumption and expenditures. Part 2. Steam, fuel oil, LPG, and all fuels

Energy Technology Data Exchange (ETDEWEB)

Patinkin, L.

1983-12-01

This report presents data on square footage and on total energy consumption and expenditures for commercial buildings in the contiguous United States. Also included are detailed consumption and expenditures tables for fuel oil or kerosene, liquid petroleum gas (LPG), and purchased steam. Commercial buildings include all nonresidential buildings with the exception of those where industrial activities occupy more of the total square footage than any other type of activity. 7 figures, 23 tables.

Back-end fuel cycle efficiencies with respect to improved uranium utilization

International Nuclear Information System (INIS)

Kuczera, B.; Hennies, H.H.

1983-01-01

The world-wide nuclear power plant (NPP) capacity is at present 160 GW(e). If one adds the power stations under construction and ordered, a plant capacity of approximately 480 GW(e) is obtained for 1990, with the share of LWRs making up more than 80%. A modern LWR consumes in the open fuel cycle about 4400 metric tonnes of natural uranium per GW(e), assuming a lifetime of 30 years and a load factor of 70%. Considering the natural uranium reserves known at present and exploitable under economic conditions, it can be conveniently estimated that, with the present NPP capacity extension perspective, the natural uranium resources may be exhausted in a few decades. This trend can be counteracted in a flexible manner by various approaches in fuel cycle technology and strategy: (i) by steady further development of the established LWR technology the uranium consumption can be reduced by about 15%; (ii) closing the nuclear fuel cycle on the basis of LWRs (i.e. thermal uranium and plutonium recycling) implies up to 40% savings in natural uranium consumption; (iii) more recent considerations include the advanced pressurized water reactor (APWR). The APWR combines the proven PWR technology with a newly developed tight lattice core with greatly improved conversion characteristics (conversion ratio = 0.90 to 0.95). In terms of uranium utilization, the APWR has an efficiency three to five times higher than a PWR; (iv) Commercial introduction of FBR systems results in an optimal utilization of uranium which, at the same time, guarantees the supply of nuclear fuel well beyond the present century. For a corresponding transition period an energy supply system can be conceived which relies essentially on extended back-end fuel cycle capacities. These would facilitate a symbiosis of PWR, APWR and FBR, characterized by high flexibility with respect to long-term developments on the energy market. (author)

Modeling of terminal-area airplane fuel consumption

Science.gov (United States)

2009-08-01

Accurate modeling of airplane fuel consumption is necessary for air transportation policy-makers to properly : adjudicate trades between competing environmental and economic demands. Existing public models used for : computing terminal-area airplane ...

Emissions and Fuel Consumption Modeling for Evaluating Environmental Effectiveness of ITS Strategies

Directory of Open Access Journals (Sweden)

Yuan-yuan Song

2013-01-01

Full Text Available Road transportation is a major fuel consumer and greenhouse gas emitter. Recently, the intelligent transportation systems (ITSs technologies, which can improve traffic flow and safety, have been developed to reduce the fuel consumption and vehicle emissions. Emission and fuel consumption estimation models play a key role in the evaluation of ITS technologies. Based on the influence analysis of driving parameters on vehicle emissions, this paper establishes a set of mesoscopic vehicle emission and fuel consumption models using the real-world vehicle operation and emission data. The results demonstrate that these models are more appropriate to evaluate the environmental effectiveness of ITS strategies with enough estimation accuracy.

Instruments for reducing the specific fuel consumption of cars

International Nuclear Information System (INIS)

Hammer, S.; Maibach, M.; Marti, P.

2001-01-01

This report for the Swiss Federal Office of Energy (SFOE) presents three possible courses of action that are to be taken to reduce the specific fuel consumption of private cars. The report first examines existing targets and the degree to which they have been met up to now, whereby the situation both in Switzerland and in the European Union is looked at. The report makes a suggestion for a future target scenario and elaborates three possible ways to met these targets: regulations on fuel-consumption, a bonus/malus system and tradable certificates. For each of the proposed instruments, the report examines implementation variants and discusses the means for their implementation. The report presents the best models for each of the implementation-variants on the basis of comparisons and the results of evaluations of their effects. For these chosen variants, the authors present comparisons of their effect on fuel consumption in graphical form and recommend tradable certificates as the best instrument

Reducing stress and fuel consumption providing road information

Directory of Open Access Journals (Sweden)

Víctor CORCOBA MAGAÑA

2014-12-01

Full Text Available In this paper, we propose a solution to reduce the stress level of the driver, minimize fuel consumption and improve safety. The system analyzes the driving style and the driver’s workload during the trip while driving. If it discovers an area where the stress increases and the driving style is not appropriate from the point of view of energy efficiency and safety for a particular driver, the location of this area is saved in a shared database. On the other hand, the implemented solution warns a particular user when approaching a region where the driving is difficult (high fuel consumption and stress using the shared database based on previous recorded knowledge of similar drivers in that area. In this case, the proposal provides an optimal deceleration profile if the vehicle speed is not adequate. Therefore, he or she may adjust the vehicle speed with both a positive impact on the driver workload and fuel consumption. The Data Envelopment Analysis algorithm is used to estimate the efficiency of driving and the driver’s workload in in each area. We employ this method because there is no preconceived form on the data in order to calculate the efficiency and stress level. A validation experiment has been conducted using both a driving simulator and a real environment with 12 participants who made 168 driving tests. The system reduced the slowdowns (38%, heart rate (4.70%, and fuel consumption (12.41% in the real environment. The proposed solution is implemented on Android mobile devices and does not require the installation of infrastructure on the road. It can be installed on any model of vehicle.

A fuel consumption model for off-road use of mobile machinery in agriculture

International Nuclear Information System (INIS)

Van linden, Veerle; Herman, Lieve

2014-01-01

Until 2009, the annual reporting of emissions by off-road transport in agriculture in Belgium was based on a 1994 calculation model that needed to be updated. An energy consumption model was established for plant production in Belgium as a backbone for a new emission model. The model starts from agricultural activities involving off-road fuel consumption. Effects of soil type, tractor size, field size and machine load are modelled. Twenty-seven FCIs (fuel consumption indicators) were computed for plant production. FCIs are expressed per year and are used for emission estimates on a regional level. FCIs ranged from 37 to 311 L/ha. Sensitivity analysis showed the highest impact of tractor size with a surplus fuel consumption between 10 and 41% depending on the crop type. Fuel consumption (L) can be further processed into greenhouse gas emissions. FCIs can be adopted in LCA (life cycle assessment) studies. With ∼310 L/ha, orchards are most fuel intensive, followed by field vegetables and sugar beets (∼150 L/ha). The total off-road energy consumption of field vegetables is high because second cropping is a common practice. - Highlights: • An energy consumption model is proposed that is based on farming activities. • No statistical data on fuel consumption are required for the calculation. • The published FCIs can be adopted in LCA studies directly or as an allocation key

Reduction of Fuel Consumption and Exhaust Pollutant Using Intelligent Transport Systems

Directory of Open Access Journals (Sweden)

Mostofa Kamal Nasir

2014-01-01

Full Text Available Greenhouse gas emitted by the transport sector around the world is a serious issue of concern. To minimize such emission the automobile engineers have been working relentlessly. Researchers have been trying hard to switch fossil fuel to alternative fuels and attempting to various driving strategies to make traffic flow smooth and to reduce traffic congestion and emission of greenhouse gas. Automobile emits a massive amount of pollutants such as Carbon Monoxide (CO, hydrocarbons (HC, carbon dioxide (CO2, particulate matter (PM, and oxides of nitrogen (NOx. Intelligent transport system (ITS technologies can be implemented to lower pollutant emissions and reduction of fuel consumption. This paper investigates the ITS techniques and technologies for the reduction of fuel consumption and minimization of the exhaust pollutant. It highlights the environmental impact of the ITS application to provide the state-of-art green solution. A case study also advocates that ITS technology reduces fuel consumption and exhaust pollutant in the urban environment.

Reduction of Fuel Consumption and Exhaust Pollutant Using Intelligent Transport Systems

Science.gov (United States)

Nasir, Mostofa Kamal; Md Noor, Rafidah; Kalam, M. A.; Masum, B. M.

2014-01-01

Greenhouse gas emitted by the transport sector around the world is a serious issue of concern. To minimize such emission the automobile engineers have been working relentlessly. Researchers have been trying hard to switch fossil fuel to alternative fuels and attempting to various driving strategies to make traffic flow smooth and to reduce traffic congestion and emission of greenhouse gas. Automobile emits a massive amount of pollutants such as Carbon Monoxide (CO), hydrocarbons (HC), carbon dioxide (CO2), particulate matter (PM), and oxides of nitrogen (NOx). Intelligent transport system (ITS) technologies can be implemented to lower pollutant emissions and reduction of fuel consumption. This paper investigates the ITS techniques and technologies for the reduction of fuel consumption and minimization of the exhaust pollutant. It highlights the environmental impact of the ITS application to provide the state-of-art green solution. A case study also advocates that ITS technology reduces fuel consumption and exhaust pollutant in the urban environment. PMID:25032239

Reduction of fuel consumption and exhaust pollutant using intelligent transport systems.

Science.gov (United States)

Nasir, Mostofa Kamal; Md Noor, Rafidah; Kalam, M A; Masum, B M

2014-01-01

Greenhouse gas emitted by the transport sector around the world is a serious issue of concern. To minimize such emission the automobile engineers have been working relentlessly. Researchers have been trying hard to switch fossil fuel to alternative fuels and attempting to various driving strategies to make traffic flow smooth and to reduce traffic congestion and emission of greenhouse gas. Automobile emits a massive amount of pollutants such as Carbon Monoxide (CO), hydrocarbons (HC), carbon dioxide (CO2), particulate matter (PM), and oxides of nitrogen (NO x ). Intelligent transport system (ITS) technologies can be implemented to lower pollutant emissions and reduction of fuel consumption. This paper investigates the ITS techniques and technologies for the reduction of fuel consumption and minimization of the exhaust pollutant. It highlights the environmental impact of the ITS application to provide the state-of-art green solution. A case study also advocates that ITS technology reduces fuel consumption and exhaust pollutant in the urban environment.

An econometrics view of worldwide fossil fuel consumption and the role of US

International Nuclear Information System (INIS)

Shafiee, Shahriar; Topal, Erkan

2008-01-01

Crude oil, coal and gas, known as fossil fuels, play a crucial role in the global economy. This paper proposes new econometrics modelling to demonstrate the trend of fossil fuels consumption. The main variables affecting consumption trends are: world reserves, the price of fossil fuels, US production and US net imports. All variables have been analysed individually for more than half a century. The research found that while the consumption of fossil fuels worldwide has increased trends in the US production and net imports have been dependent on the type of fossil fuels. Most of the US coal and gas production has been for domestic use, which is why it does not have a strong influence on worldwide fossil fuel prices. Moreover, the reserves of fossil fuels have not shown any diminution during the last couple of decades and predictions that they were about to run out are not substantiated. The nominal and real price of fossil fuels was found to change depending on the type. Finally, estimates of three econometric models for the consumption of fossil fuels from 1949 to 2006 are presented which identify the effects of significant variables

Domestic fuel consumption in Estonia (The first half of 1993)

International Nuclear Information System (INIS)

Veski, Rein

1993-01-01

According to the data of the Estonian Statistics Department the consumption of fuels in Estonia over the period of January-June 1993 was a follows: coal 77,053 t, oil shale 7,981,840 t, fuel peat 127,376 t, peat briquette 42,041 t, firewood 154,511 solid metres, wood chip 55,648 solid metres. The consumption of fuel peat (chiefly for producing briquette) was more considerable in the town of Kohtla-Jaerve and in the districts of Paernu and Tartu where Estonia's major peat-briquette plants are located. The share of peat in the other industries has been less significant. (author)

Clarification of dissolved irradiated light-water-reactor fuel

International Nuclear Information System (INIS)

Rodrigues, G.C.

1983-02-01

Bench-scale studies with actual dissolved irradiated light water reactor (LWR) fuels showed that continuous centrifugation is a practical clarification method for reprocessing. Dissolved irradiated LWR fuel was satisfactorily clarified in a bench-scale, continuous-flow bowl centrifuge. The solids separated were successfully reslurried in water. When the reslurried solids were mixed with clarified centrate, the resulting suspension behaved similar to the original dissolver solution during centrifugation. Settling rates for solids in actual irradiated fuel solutions were measured in a bottle centrifuge. The results indicate that dissolver solutions may be clarified under conditions achievable by available plant-scale centrifuge technology. The effective particle diameter of residual solids was calculated to be 0.064 microns for Oconee-1 fuel and 0.138 microns for Dresden-1 fuel. Filtration was shown unsuitable for clarification of LWR fuel solutions. Conventional filtration with filter aid would unacceptably complicate remote canyon operation and maintenance, might introduce dissolved silica from filter aids, and might irreversibly plug the filter with dissolver solids. Inertial filtration exhibited irreversible pluggage with nonradioactive stand-in suspensions under all conditions tested

Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

International Nuclear Information System (INIS)

Olson, W.O.; Nigg, D.W.

1981-05-01

The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

Analysis and synthesis of the theoretical studies performed on the control and safety of LWR's burning plutonium fuel

International Nuclear Information System (INIS)

Basselier, J.; Renard, A.; Holzer, R.; Hnilica, K.

1982-01-01

This report presents the comparative investigations of parameters for plutonium fuelled power stations (PWR and BWR) under steady state and dynamic conditions for typical accidents. The recycling of about 30% of mixed oxide fuel in the large LWR cores should not induce special problems, if some cautions are taken in core design to minimize the differences with UO 2 cores taking into account a limited margin fo uncertainty. The influence on the core behaviour, during the investigated accidents, is not very important and does not induce restrictions for at least a 30% Pu fraction in the core. The operation with high plutonium amounts may be considered. From the steady state and safety point-of-views, the maximum allowable quantity into the cores should be sought for each reactor. In principle, a 100% UO 2 -PuO 2 core could be operated under certain conditions of loading pattern and shutdown margins. For what concerns the storage and handling, the studies show the following results: storage pool design with respect to criticality will not be affected by the use of UO 2 -PuO 2 fuel asemblies

Engine thermomanagement with electrical components for fuel consumption reduction

Energy Technology Data Exchange (ETDEWEB)

Cortona, E.; Onder, C.H.; Guzzella, L. [Swiss Federal Inst. of Technology, Zurich (Switzerland)

2002-09-01

This paper proposes a solution for advanced temperature control of the relevant temperature of a combustion engine. It analyses the possibility of reducing vehicle fuel consumption by improving engine thermomanagement. In conventional applications, combustion engine cooling systems are designed to guarantee sufficient heat removal at full load. The cooling pump is belt-driven by the combustion engine crankshaft, resulting in a direct coupling of engine and cooling pump speeds. It is dimensioned such that it can guarantee adequate performance over the full engine speed range. This causes an excessive flow of cooling fluid at part-load conditions and at engine cold-start. This negatively affects the engine efficiency and, as a consequence, the overall fuel consumption. Moreover, state-of-the-art cooling systems allow the control of the coolant temperature only by expansion thermostats (solid-to-liquid phase wax actuators). The resulting coolant temperature does not permit engine efficiency to be optimized. In this paper, active control of the coolant flow as well as of the coolant temperature has been realized using an electrical cooling pump and an electrically driven valve which controls the flow distribution between the radiator and its bypass. For this purpose, a control-oriented model of the whole cooling system has been derived. Model-based feedforward and feedback controls of coolant temperature and flow have been designed and tested. With the additional actuators and the model-based control scheme, a good performance in terms of fast heat-up and small temperature overshoot has been achieved. The improvements in fuel consumption obtained with the proposed configuration have been verified on a dynamic testbench. Both engine cold-start under stationary engine operation and the European driving cycle MVEG-A with engine cold-start were tested. The fuel consumption reductions achieved during these tests vary between 2.8 and 4.5 per cent, depending on the engine

Nondestructive verification and assay systems for spent fuels

International Nuclear Information System (INIS)

Cobb, D.D.; Phillips, J.R.; Bosler, G.E.; Eccleston, G.W.; Halbig, J.K.; Hatcher, C.R.; Hsue, S.T.

1982-04-01

This is an interim report of a study concerning the potential application of nondestructive measurements on irradiated light-water-reactor (LWR) fuels at spent-fuel storage facilities. It describes nondestructive measurement techniques and instruments that can provide useful data for more effective in-plant nuclear materials management, better safeguards and criticality safety, and more efficient storage of spent LWR fuel. In particular, several nondestructive measurement devices are already available so that utilities can implement new fuel-management and storage technologies for better use of existing spent-fuel storage capacity. The design of an engineered prototype in-plant spent-fuel measurement system is approx. 80% complete. This system would support improved spent-fuel storage and also efficient fissile recovery if spent-fuel reprocessing becomes a reality

Spent fuel data base: commercial light water reactors

International Nuclear Information System (INIS)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel

Model for Estimation of Fuel Consumption of Cruise Ships

Directory of Open Access Journals (Sweden)

Morten Simonsen

2018-04-01

Full Text Available This article presents a model to estimate the energy use and fuel consumption of cruise ships that sail Norwegian waters. Automatic identification system (AIS data and technical information about cruise ships provided input to the model, including service speed, total power, and number of engines. The model was tested against real-world data obtained from a small cruise vessel and both a medium and large cruise ship. It is sensitive to speed and the corresponding engine load profile of the ship. A crucial determinate for total fuel consumption is also associated with hotel functions, which can make a large contribution to the overall energy use of cruise ships. Real-world data fits the model best when ship speed is 70–75% of service speed. With decreased or increased speed, the model tends to diverge from real-world observations. The model gives a proxy for calculation of fuel consumption associated with cruise ships that sail to Norwegian waters and can be used to estimate greenhouse gas emissions and to evaluate energy reduction strategies for cruise ships.

Synergistic fuel cycles of the future

International Nuclear Information System (INIS)

Meneley, D.A.; Dastur, A.R.

1997-01-01

Good neutron economy is the basis of the fuel cycle flexibility in the CANDU reactor. This paper describes the fuel cycle options available to the CANDU owner with special emphasis on resource conservation and waste management. CANDU fuel cycles with low initial fissile content operate with relatively high conversion ratio. The natural uranium cycle provides over 55 % of energy from the plutonium that is created during fuel life. Resource utilization is over 7 MWd/kg NU. This can be improved by slight enrichment (between 0.9 and 1.2 wt % U235) of the fuel. Resource utilization increases to 11 MWd/kg NU with the Slightly Enriched Uranium cycle. Thorium based cycles in CANDU operate at near-breeder efficiency. Obey provide attractive options when used with natural uranium or separated (reactor grade and weapons grade) plutonium as driver fuels. In the latter case, the energy from the U233 plus the initial plutonium content amounts to 3.4 GW(th).d/kg Pu-fissile. The same utilization is expected from the use of FBR plutonium in a CANDU thorium cycle. Extension of natural resource is achieved by the use of spent fuels in CANDU. The LWR/CANDU Tandem cycle leads to an additional 77 % of energy through the use of reprocessed LWR fuel (which has a fissile content of 1.6 wt %) in CANDU. Dry reprocessing of LWR fuel with the OREOX process (a more safeguardable alternative to the PUREX process) provides an additional 50 % energy. Uranium recovered (RU) from separation of plutonium contained in spent LWR fuel provides an additional 15 MWd/kg RU. CANDU's low fissile requirement provides the possibility, through the use of non-fertile targets, of extracting energy from the minor actinides contained in spent fuel. In addition to the resource utilization advantage described above, there is a corresponding reduction in waste arisings with such cycles. This is especially significant when separated plutonium is available as a fissile resource. (author)

Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-12-01

At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

Canadian power reactor fuel

International Nuclear Information System (INIS)

Page, R.D.

1976-03-01

The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

Uranium plutonium oxide fuels

International Nuclear Information System (INIS)

Cox, C.M.; Leggett, R.D.; Weber, E.T.

1981-01-01

Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

Projection of fossil fuels consumption in the Venezuelan electricity generation industry

International Nuclear Information System (INIS)

Vidoza, Jorge A.; Gallo, Waldyr L.R.

2016-01-01

This study presents a prospective analysis on the impacts of recent efficient energy policies application in Venezuela, integrating both oil production and electricity supply to assess energy resources balance in a quantitative manner. A special focus is given to main fossil fuels used in the electric power industry; natural gas, diesel oil and fuel oil. Four scenarios were proposed, ranging from a low-economy-growth/low-efficiency scenario to an optimist high-economy-growth/high-efficiency scenario. Efficiency effects are more notorious for high-economy-growth case, fuel consumption for electricity generation reduces 38% for natural gas, 12% for diesel and 29% for fuel oil, in the established time period. Deficits in oil and gas Venezuelan production were also determined, deficits are highly affected by economical forecasting, and by fuel smuggling in Venezuelan borders. Results showed the high importance of energy efficiency policies development for Venezuela, in order to reduce fossil fuel domestic consumption to allocate them in a more profitable market. - Highlights: • We made a prospective analysis on efficient energy policies impacts in Venezuela. • Reduced fuel consumption was obtained for efficient scenarios. • Current energy regulations are not enough to encourage energy efficiency. • Hydroelectricity projects need more promotion to have deeper impacts.

Fuel Consumption Management in the Transportation Sector in Iran

DEFF Research Database (Denmark)

Dastjerdi, Aliasghar M.; Araghi, Bahar Namaki

2011-01-01

Energy consumption in the transportation sector in Iran is significantly higher than global norms and standards which caused some issues including wasting national resources, deteriorating air quality, GHG emissions etc. The major purpose of this paper is to introduce practical policies, strategies...... and technologies to reduce liquid fuel consumption known as a dominant source of energy in transport sector in Iran. Since, the road subsector has the major share in consuming liquid fuel amongst others, more attention is given to the methods for reducing consumption in this subsector. The relating policies...... and actions were classified by optimization measures according to four separate categories as follows; “Optimization of Supply of Transportation Services”, “Optimization of Transport Demand”, “Optimization of Energy Consumption” and “Optimization of Car Manufacturing”....

Reactor Structure Materials: Nuclear Fuel

International Nuclear Information System (INIS)

Sannen, L.; Verwerft, M.

2000-01-01

Progress and achievements in 1999 in SCK-CEN's programme on applied and fundamental nuclear fuel research in 1999 are reported. Particular emphasis is on thermochemical fuel research, the modelling of fission gas release in LWR fuel as well as on integral experiments

Global exergetic dimension of hydrogen use in reducing fossil fuel consumption

International Nuclear Information System (INIS)

Adnan Midilli; Ibrahim Dincer

2009-01-01

In this paper, hydrogen is considered as a renewable and sustainable solution for minimizing the fossil fuel based-global irreversibility coefficient of global fossil fuel consumption and combating global warming and studied exergetically through a parametric performance analysis. The environmental impact results are then compared with the ones obtained for fossil fuels. In this regard, some exergetic expressions such as global waste exergy factor, global irreversibility coefficient and hydrogen based-global exergetic indicator. In order to investigate the role of hydrogen use at minimizing the fossil fuel based global irreversibility, the actual fossil fuel consumption data are taken from the literature. Due to the unavailability of appropriate hydrogen data for analysis, it is assumed that the utilization ratios of hydrogen are ranged between 0 and 1. Consequently, if exergetic utilization ratio of hydrogen from non-fossil fuel sources at a certain exergetic utilization ratio of fossil fuels increases, the fossil fuel based-global irreversibility coefficient will decrease. (author)

Development of Advanced High Uranium Density Fuels for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

2016-02-15

This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

Biomass burning fuel consumption rates: a field measurement database

NARCIS (Netherlands)

van Leeuwen, T.T.; van der Werf, G.R.; Hoffmann, A.A.; Detmers, R.G.; Ruecker, G.; French, N.H.F.; Archibald, S.; Carvalho Jr., J.A.; Cook, G.D.; de Groot, J.W.; Hely, C.; Kasischke, E.S.; Kloster, S.; McCarty, J.L.; Pettinari, M.L.; Savadogo, P.

2014-01-01

Landscape fires show large variability in the amount of biomass or fuel consumed per unit area burned. Fuel consumption (FC) depends on the biomass available to burn and the fraction of the biomass that is actually combusted, and can be combined with estimates of area burned to assess emissions.

Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

Energy Technology Data Exchange (ETDEWEB)

Brueziere, J.; Tribout-Maurizi, A.; Durand, L.; Bertrand, N. [Recycling Business Unit, AREVA, 1 place de la coupole, 92084 Paris La defense Cedex (France)

2013-07-01

Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy

International Nuclear Information System (INIS)

Fukaya, Y.; Goto, M.

2017-01-01

Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR

Fuel Consumption and Emissions from Airport Taxi Operations

Science.gov (United States)

Jung, Yoon

2010-01-01

Developed a method to calculate fuel consumption and emissions of phases of taxi operations. Results at DFW showed that up to 18% of fuel can be saved by eliminating stop-and-go situations. Developed an energy efficient and environmentally friendly surface concept: Spot and Runway Departure Advisory (SARDA) tool. The SARDA tool has been identified as a potential candidate for a technology transfer to the FAA.

Fuel consumption reduction in urban buses by using power split transmissions

International Nuclear Information System (INIS)

Macor, Alarico; Rossetti, Antonio

2013-01-01

Highlights: • Fuel consumption is highly influenced by transmission layout and control. • Continuous Variable Transmission allows control of the working point of the engine. • Hydromechanical CVT was numerically tested and compared with usual mechanical solutions. • CVT solution proved to reduce fuel consumption, improving the I.C.E. mean efficiency. • Results obtained from standard emission test cycles: New York and Manhattan Buses. - Abstract: The reduction of fuel consumption of urban vehicles is one of the challenges of the society today. Many promising technologies such as hybrid electric vehicles and fuel cells are the focus of a lot of research and investments, but high costs prevent their development on a large scale, especially in heavy duty vehicles. Simpler solutions, such as continuously variable transmissions, are then interesting alternatives: they preserve some of the advantages of hybrid solutions significantly reducing the overall cost of the system. Since these technologies allow the decoupling of the engine’s velocity from that of the wheels, optimised management of the engine becomes feasible, leading to lower fuel consumption than that of traditional solutions. This work compares the hydromechanical transmission with a traditional power shift in an application where efficiency and comfort are required: a city bus. The performance of a vehicle with a dual stage hydromechanical transmission has been investigated by means of a model implemented in the AMESim environment. A second model of a vehicle with a power shift transmission has been taken as reference solution. During the simulation of two standard test cycles, the first vehicle showed the lower fuel consumption, due to its better engine management, and to the operation of the torque converter during starts, which lowers the efficiency of the reference vehicle

Modeling transit bus fuel consumption on the basis of cycle properties.

Science.gov (United States)

Delgado, Oscar F; Clark, Nigel N; Thompson, Gregory J

2011-04-01

A method exists to predict heavy-duty vehicle fuel economy and emissions over an "unseen" cycle or during unseen on-road activity on the basis of fuel consumption and emissions data from measured chassis dynamometer test cycles and properties (statistical parameters) of those cycles. No regression is required for the method, which relies solely on the linear association of vehicle performance with cycle properties. This method has been advanced and examined using previously published heavy-duty truck data gathered using the West Virginia University heavy-duty chassis dynamometer with the trucks exercised over limited test cycles. In this study, data were available from a Washington Metropolitan Area Transit Authority emission testing program conducted in 2006. Chassis dynamometer data from two conventional diesel buses, two compressed natural gas buses, and one hybrid diesel bus were evaluated using an expanded driving cycle set of 16 or 17 different driving cycles. Cycle properties and vehicle fuel consumption measurements from three baseline cycles were selected to generate a linear model and then to predict unseen fuel consumption over the remaining 13 or 14 cycles. Average velocity, average positive acceleration, and number of stops per distance were found to be the desired cycle properties for use in the model. The methodology allowed for the prediction of fuel consumption with an average error of 8.5% from vehicles operating on a diverse set of chassis dynamometer cycles on the basis of relatively few experimental measurements. It was found that the data used for prediction should be acquired from a set that must include an idle cycle along with a relatively slow transient cycle and a relatively high speed cycle. The method was also applied to oxides of nitrogen prediction and was found to have less predictive capability than for fuel consumption with an average error of 20.4%.

Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

Science.gov (United States)

Snead, L. L.; Terrani, K. A.; Katoh, Y.; Silva, C.; Leonard, K. J.; Perez-Bergquist, A. G.

2014-05-01

This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle-matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320-360 °C range to a maximum dose of 7.7 × 1025 n/m2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix.

Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

International Nuclear Information System (INIS)

Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

1978-10-01

This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

International Nuclear Information System (INIS)

Pope, Michael A.; Sen, R. Sonat; Boer, Brian; Ougouag, Abderrafi M.; Youinou, Gilles

2011-01-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

Economics of spent LWR fuel storage

International Nuclear Information System (INIS)

Clark, H.J.

1980-01-01

A low cost option for spent fuel inventories would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) storage facilities, and away-from-reactor (AFR) storage facilities. Fuel storage requirements will be met best by transfer of fuel or by re-racking existing reactor basins whenever these options are available. These alternatives represent not only the lowest cost storage options but also the most timely. Fuel can be shipped to other storage pools for about $10/kg depending on the distance, while costs for reracking range from $18 to 25/kg depending on the approach. These alternatives are recognized to face environmental and regulatory obstacles. However, such obstacles should be less severe than similar issues that would be encountered with AR or AFR basin storage. When storage requirements cannot be met by the first two options, the next least costly alternative for most utilities will be use of a Federal AFR. Storage cost of about $137/kg at an AFR are less costly than charges of up to $350/kg that could be incurred by the use of AR basins. AR basins are practical only when a utility requires storage capacity to accommodate annual additions of 100 MT or more of spent fuel. The large reactor complexes discharging this much feul are not currently those that require relief from fuel storage problems. A recent development in Germany may offer an AR alternative of dry storage in transportation/storage casks at a cost of $200/kg; however, this method has not yet been accepted and licensed for use in the US

Investigation of Factors Affecting Fuel Consumption of Rural Households in Central District of Zahedan County

Directory of Open Access Journals (Sweden)

Maryam Sharifzadeh

2014-08-01

Full Text Available Domestic fuel forms a significant part of the total energy demand and providing adequate and sustainable fuel is considered as a pivotal foundation for development. This survey study aims at investigating domestic fuel consumption patterns of rural areas in central district of Zahedan County, Eastern Iran. The sample was consisted of 250 household heads resided in rural areas with more than 25 households. The validity of questionnaire was confirmed by a panel of experts and its reliability was measured by using a pilot study. Findings revealed that, only 43 percent (n=102 household revealed an efficient energy use behavior. The efficient energy consumers were significantly different with respect to their education level, age, income, and the other household characteristics. Results from the structural equation modeling which was used to confirm adequacy of the reasoned action model of fuel consumption behavior, showed that 0.24 percent of fuel consumption behavior was determined by attitude, intention and subjective norms towards fuel consumption. The paper presented applied suggestions regarding fuel consumption with special consideration on health and ecosystem and indoor sanitation issues.

Nuclear fuel

International Nuclear Information System (INIS)

D Hondt, P.

1998-01-01

The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

SPES, Fuel Cycle Optimization for LWR

International Nuclear Information System (INIS)

1973-01-01

1 - Nature of physical problem solved: Determination of optimal fuel cycle at equilibrium for a light water reactor taking into account batch size, fuel enrichment, de-rating, shutdown time, cost of replacement energy. 2 - Method of solution: Iterative method

Transition period fuel cycle from current to next generation reactors for Japan

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2007-01-01

Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named Flexible Fuel Cycle Initiative (FFCI)' which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual 'Recycle Material' is, according to FBRs introduction status, immediately treated in the FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles. (author)

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

International Nuclear Information System (INIS)

Johnson, E.R.; Best, R.E.

2009-01-01

produced and consumed in a fleet of 100 1,000 MWe LWRs and in FRs. The model also included recycle and reuse of extant inventories of spent LWR fuel. The results of the simulations allowed comparisons of the two fuel cycles from the standpoints of cost, non-proliferation, radiological health, wastes generated, and sustainability. Results of the research also provide insights regarding (i) multiple recycling of uranium and plutonium from spent MOX fuel in LWRs, (ii) costs and benefits of reenriching and reusing uranium from spent LWR fuel; (iii) effects of using uranium, plutonium, and minor actinides from LWR spent fuels to produce fuel for FRs; (iv) net rates of consumption (burning) in FRs of actinide elements produced in LWRs, and (v) ependencies of and interactions among the different systems of an advanced nuclear fuel cycle -- and the flows of nuclear materials between these systems.

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

Energy Technology Data Exchange (ETDEWEB)

E. R. Johnson; R. E. Best

2009-12-28

materials were produced and consumed in a fleet of 100 1,000 MWe LWRs and in FRs. The model also included recycle and reuse of extant inventories of spent LWR fuel. The results of the simulations allowed comparisons of the two fuel cycles from the standpoints of cost, non-proliferation, radiological health, wastes generated, and sustainability. Results of the research also provide insights regarding (i) multiple recycling of uranium and plutonium from spent MOX fuel in LWRs, (ii) costs and benefits of reenriching and reusing uranium from spent LWR fuel; (iii) effects of using uranium, plutonium, and minor actinides from LWR spent fuels to produce fuel for FRs; (iv) net rates of consumption (burning) in FRs of actinide elements produced in LWRs, and (v) ependencies of and interactions among the different systems of an advanced nuclear fuel cycle -- and the flows of nuclear materials between these systems.

Economics of spent LWR fuel storage

International Nuclear Information System (INIS)

Clark, H.J.; O'Neill, G.F.

1980-01-01

A power reactor operator, confronted with rising spent fuel inventories that would soon exceed his storage capacity, has to decide what to do with this fuel if he wants to continue reactor operations. A low cost option would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) basins for storage, and away-from-reactor (AFR) basins for storage. Economic considerations for each of the alternatives are compared

The effect of conservation tillage forward speed and depth on farm fuel consumption

Directory of Open Access Journals (Sweden)

A Jalali

2015-09-01

Full Text Available Introduction: In recent years, production techniques and equipment have been developed for conservation tillage systems that have been adopted by many farmers. With proper management, overall yield averages for conventional and reduced tillage systems are nearly identical. Sometimes, field operations can be combined by connecting two or more implements. Much research has focused on either reducing or eliminating tillage operations to develop sustainable crop production methods. The greatest costs in farm operations are associated with tillage due to greater specific energy requirement in tillage and the high fuel costs. Combined operations reduce both fuel consumption and time and labor requirements by eliminating at least one individual trip over the field. Light tillage, spraying, or fertilizing operations can be combined with eitherprimary or secondary tillage or planting operations. The amount of fuel saved depends on the combined operations. Generally, light tillage, spraying, and fertilizing operations consume between 0.25 and 0.50 gallons of diesel fuel per acre. Fuel savings of 0.12 to 0.33 gallons per acre can usually be expected from combining operations. Eliminating one primary tillage operation and combining one light tillage, spraying, or fertilizing operation with another tillage or planting operation can usually save at least a gallon of diesel fuel per acre. Combining operations has the added benefit of reducing wheel traffic and compaction. To improve the tillage energy efficiency, implementing effective and agronomic strategies should be improved. Different tillage systems should be tested to determine the most energy efficient ones. Tillage helps seed growth and germination through providing appropriate conditions for soil to absorb sufficient temperature and humidity. Tillage is a time consuming and expensive procedure. With the application of agricultural operations, we can save considerable amounts of fuel, time and

Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

1995-06-01

This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

Preliminary study on direct recycling of spent PWR fuel in PWR system

International Nuclear Information System (INIS)

Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

2012-01-01

Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

The applicability of detailed process for neutron resonance absorption to neutronics analyses in LWR next generation fuels to extend burnup

International Nuclear Information System (INIS)

Kameyama, Takanori; Nauchi, Yasushi

2004-01-01

Neutronics analyses with detail processing for neutron resonance absorption in LWR next generation UOX and MOX fuels to extend burnup were performed based on the neutronic transport and burnup calculation. In the detailed processing, ultra-fine energy nuclear library and collision probabilities between neutron and U, Pu nuclides (actinide nuclides) are utilized for two-dimension geometry. In the usual simple processing (narrow resonance approximation), shielding factors and compensation equations for neutron resonance absorption are utilized. The results with detailed and simple processing were compared to clarify where the detailed processing is needed. The two processing caused difference of neutron multiplication factor by 0.5% at the beginning of irradiation, while the difference became smaller as burnup increased and was not significant at high burnup. The nuclide compositions of the fuel rods for main actinide nuclides were little different besides Cm isotopes by the processing, since the neutron absorption rate of 244 Cm became different. The detail processing is needed to evaluate the neutron emission rate in spent fuels. In the fuel assemblies, the distributions of rod power rates were not different within 0.5%, and the peak rates of fuel rod were almost the same by the two processing at the beginning of irradiation when the peak rate is the largest during the irradiation. The simple processing is also satisfied for safety evaluation based on the peak rate of rod power. The difference of local power densities in fuel pellets became larger as burnup increased, since the neutron absorption rate of 238 U in the peripheral region of pellets were significantly different by the two processing. The detail processing is needed to evaluate the fuel behavior at high burnup. (author)

Microencapsulated fuel technology for commercial light water and advanced reactor application

International Nuclear Information System (INIS)

Terrani, Kurt A.; Snead, Lance L.; Gehin, Jess C.

2012-01-01

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

Alternative transport fuels: supply, consumption and conservation

International Nuclear Information System (INIS)

Trindade, S.C.

1990-01-01

Road-based passenger and freight transport almost exclusively uses petroleum/hydrocarbon fuels in the fluid form. These fuels will probably continue to be major transport fuels well into the 21st century. As such there is need to prolong their use which can be done through: (1) conservation of fuel by increasing efficiency of internal combustion engines, and (2) conversion of natural gas, coal and peat, and biomass into alternate fuels such as ethanol, methanol, CNG, LNG, LPG, low heat-content (producer) gas and vegetable oils. Research, development and demonstration (RD and D) priorities in supply, consumption and conservation of these alternate fuels are identified and ranked in the context of situation prevailing in Brazil. Author has assigned the highest priority for research in the impact of pricing, economic, fiscal and trade policies, capital allocation criteria and institutional and legislative framework. It has also been emphasised that an integrated or systems approach is mandatory to achieve net energy gains in transport sector. (M.G.B.). 33 refs., 11 tabs., 4 figs

Study on a method for estimating fuel consumption in a seaway

DEFF Research Database (Denmark)

Iseki, Toshio; Nielsen, Ulrik Dam

2013-01-01

direction has a great influence on the main engine horse power and fuel consumption, and also shows a possibility of fuel efficiency prediction. In order to develop an eco-friendly navigation support system, results of Bayesian wave estimation are applied to fuel efficiency prediction. The Bayesian method...

Fuel-consumption reflects increasing export

International Nuclear Information System (INIS)

TREND

2003-01-01

Fuel consumption in Slovakia has been substantially rising during the first three months of this year. During January and February about 6 percent more petrol and 15 percent more diesel fuel were sold compared to last year. Retail sales were growing even faster, informed president of Slovak Association of Petroleum Industry and Trade (SAPPO), Karoly Robak. He considers this an evidence of a substantial economic growth. The export is executed in form of freight transport (lorries), as this, together with agriculture, is the largest diesel fuel consumer. K. Robak, who at the same time is the Deputy General Manager For Retail at Slovnaft, a.s., Bratislava, expects the sales of fuel to grow by about 8 percent this year. Low prices due to a weaker US dollar should only support these tendencies. Last year the sales grew by 8,3 percent and in retail business the increase exceeded 10 percent. At the moment there are about 700 petrol stations in Slovakia (SAPPO) and over 330 out of them are owned by Slovnaft. The sale of fuel is now moving more to the retail business - to petrol stations as the price difference between retail and wholesale is decreasing. Last year for instance the wholesale recorded an increase in sales of diesel by 6,5 percent whereas retail sales grew by 17 percent. With petrol the situation was similar - 6,4 increase in retail sales and 2,5 percent in wholesale. (Author)

Effect of Casmo-5 cross-section data and doppler temperature definitions on LWR reactivity initiated accidents - 166

International Nuclear Information System (INIS)

Grandi, G.; Smith, K.; Xu, Z.; Rhodes, J.

2010-01-01

During LWR Reactivity Initiated Accidents (RIA), the accurate evaluation of the Doppler reactivity feedback depends on the Doppler coefficient computed by the lattice physics code (e.g. CASMO-5), and on the effective Doppler temperature computed by the transient code (e.g. SIMULATE-3K) using the non-uniform intra-pellet temperature profile. CASMO-5 has many new features compared with its predecessor. Among them, the replacement of the L-library (based primarily on ENDF/B IV data) by the latest available nuclear data (ENDF/B VII.0), and the Monte Carlo based resonance elastic scattering model to overcome deficiencies in NJOY modeling have a significant impact on the fuel temperature coefficient, and hence on LWR RIA. The Doppler temperature effect in thermal reactors is driven by the 238 U absorption. The different effective Doppler temperature definitions, available in the literature, try to capture the considerable self-shielding of the 238 U absorption that occurs in the pellet surface by defining an appropriate fuel temperature to compute cross-sections. In this work, we investigate the effect of the nuclear data generated by CASMO-5 on RIA, as well as the impact of different effective Doppler temperature definitions, including one proposed by the authors. It is concluded: 1) LWR RIA evaluated using CASMO-5 cross section data will be milder because the energy released is ∼10% smaller; 2) the prompt enthalpy rise is barely affected by the choice of the Doppler temperature definition; and 3) the peak fuel enthalpy is affected by the choice of the Doppler temperature definition, the under-prediction of the Doppler reactivity by the 'NEA' Doppler temperature results in a conservative estimate of the peak fuel enthalpy. (authors)

Spent fuel data base: commercial light water reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Reconfiguration of photovoltaic panels for reducing the hydrogen consumption in fuel cells of hybrid systems

Directory of Open Access Journals (Sweden)

Daniel González-Montoya

2017-05-01

Full Text Available Hybrid generation combines advantages from fuel cell systems with non-predictable generation approaches, such as photovoltaic and wind generators. In such hybrid systems, it is desirable to minimize as much as possible the fuel consumption, for the sake of reducing costs and increasing the system autonomy. This paper proposes an optimization algorithm, referred to as population-based incremental learning, in order to maximize the produced power of a photovoltaic generator. This maximization reduces the fuel consumption in the hybrid aggregation. Moreover, the algorithm's speed enables the real-time computation of the best configuration for the photovoltaic system, which also optimizes the fuel consumption in the complementary fuel cell system. Finally, a system experimental validation is presented considering 6 photovoltaic modules and a NEXA 1.2KW fuel cell. Such a validation demonstrates the effectiveness of the proposed algorithm to reduce the hydrogen consumption in these hybrid systems.

Motorcycle On-Road Driving Parameters Influencing Fuel Consumption and Emissions on Congested Signalized Urban Corridor

Directory of Open Access Journals (Sweden)

Atthapol Seedam

2017-01-01

Full Text Available This study aims to find the on-road driving parameters influencing fuel consumption and emissions of motorcycle driving on a congested signalized urban corridor. A motorcycle onboard measurement system was developed to measure instantaneously and continuously record on-road driving data, including speed-time profile, emissions, and fuel consumption, by the second. The test motorcycles were driven by 30 sample motorcyclists on a signalized urban corridor in Khon Kaen City, Thailand, to collect their on-road driving behavior during the morning peak period. Cluster analysis was applied to analyze collected driving data and to categorize the drivers by level of fuel consumption and on-road driver behavior. The on-road driving parameter influencing fuel consumption and emissions was then determined. Results revealed that proportion of idle time significantly influenced fuel consumption and emissions of motorcycle driving on a congested signalized urban corridor, though aggressive driving behavior, hard acceleration and deceleration, did not have the same kind of influence.

Fuel cycle integration issues associated with P/T technology

International Nuclear Information System (INIS)

Michaels, G.E.; Ludwig, S.B.

1992-01-01

The three primary interfaces between a generic partitioning and transmutation (P/T) technology and the existing United States fuel cycle are the light-water reactor (LWR) spent fuel inventory, the reprocessed uranium (RU) stream, and the high-level waste stream. The features and implications of these three interfaces are reviewed and their implications for P/T system design and for waste management are assessed. The variability of transuranic nuclide composition in the LWR spent fuel is calculated and its potential implications for transmutation system core design are discussed. The radiological characteristics of the RU stream are presented, and options for disposition of the stream are reviewed. Most P/T scenarios assume that RU will be recycled to LWRs. This study demonstrates, however, that LWR recycle cannot totally consume the reprocessed stream, and disposal of a waste uranium steam with high levels of radiologically-significant isotopes will still be necessary. The radioactivity of the tails stream for enrichment plants resulting from a dedicated RU campaign is calculated. The tendency of gaseous diffusion plant enrichment technology to deplete the tails stream of minor uranium isotopes is seen as a benefit and an advantage over Atomic Vapor Laser Isotope Separation-type technology. Finally, the implications of P/T on LWR-origin wastes reporting to the repository is discussed, and several significant differences between LWR-origin waste originating from transmutation systems are assessed

Minimizing fuel wood consumption through the evolution of hot ston ...

African Journals Online (AJOL)

The central objective of this paper is to minimize fuelwood consumption through evolving alternative domestic energy. Data on alternative domestic energy sources, and use fuel wood consumption during scarcity of petroleum were collected using structured questionnaires. Data on time spent to cook yam, race and beans ...

Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

International Nuclear Information System (INIS)

Matsumura, Tetsuo

1994-01-01

Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

Proceedings of fuel safety research specialists' meeting

International Nuclear Information System (INIS)

Suzuki, Motoe

2002-08-01

Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's ''Advanced LWR Fuel Performance and Safety Research Program''. A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting. The 10 of the presented papers are indexed individually. (J.P.N.)

Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

Energy Technology Data Exchange (ETDEWEB)

1980-03-25

The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options.

Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

International Nuclear Information System (INIS)

1980-01-01

The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options

Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

International Nuclear Information System (INIS)

Williams, Kent Alan; Shropshire, David E.

2009-01-01

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

Roof Box Shape Streamline Adaptation and the Impact towards Fuel Consumption

Directory of Open Access Journals (Sweden)

Abdul Latif M.F.

2017-01-01

Full Text Available The fuel price hike is currently a sensational national issue in Malaysia. Since the rationalization of fuel subsidies many were affected especially the middle income family. Vehicle aerodynamic were directly related to the fuel consumption, were extra frontal area result a higher drag force hence higher fuel consumption. Roof box were among the largest contributor to the extra drag, thus the roof box shape rationalization were prominent to reduce the extra drag. The idea of adopting water drop shape to the roof box design shows prominent result. The roof box has been simulated using MIRA virtual wind tunnel modelling via commercial computational fluid dynamic (CFD package. This streamline shape drastically reduce the drag force by 34% resulting to a 1.7% fuel saving compare to the conventional boxy roof box. This is an effort to reduce the carbon foot print for a sustainable green world.

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2E, Physical descriptions of LWR nonfuel assembly hardware, Appendix 2F, User's guide to the LWR nonfuel assembly data base

International Nuclear Information System (INIS)

1987-12-01

This appendix includes a two to three page Physical Description report for each Non-fuel Assembly (NFA) Hardware item identified from the current data. Information was obtained via subcontracts with these NFA hardware vendors: Babcock and Wildox, Combustion Engineering and Westinghouse. Data for some NFA hardware are not available. For such hardware, the information shown in this report was obtained from the open literature. Efforts to obtain additional information are continuing. NFA hardware can be grouped into six categories: BWR Channels, Control Elements, Guide Tube Plugs/Orifice Rods, Instrumentation, Neutron Poisons, and Neutron Sources. This appendix lists Physical Description reports alphabetically by vendor within each category. Individual Physical Description reports can be generated interactively through the menu-driven LWR Non-Fuel Assembly Hardware Data Base system. These reports can be viewed on the screen, directed to a printer, or saved in a text file for later use. Special reports and compilations of specific data items can be produced on request

Fuel consumption and associated emissions from seagoing ships at berth derived from an on-board survey

NARCIS (Netherlands)

Hulskotte, J.H.J.; Denier van der Gon, H.A.C.

2010-01-01

A methodology is presented to estimate the emissions of ships at berth based on their actual fuel consumption and the fuel quality. Accurate estimates of emissions from ships at berth demand reliable knowledge of the fuel consumption while at berth and associated fuel characteristics. However,

Fuel cycle

International Nuclear Information System (INIS)

Bahm, W.

1989-01-01

The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

International Nuclear Information System (INIS)

Snead, L.L.; Terrani, K.A.; Katoh, Y.; Silva, C.; Leonard, K.J.; Perez-Bergquist, A.G.

2014-01-01

This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle–matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320–360 °C range to a maximum dose of 7.7 × 10 25 n/m 2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix

Light water reactor fuel reprocessing and recycling

International Nuclear Information System (INIS)

1977-07-01

This document was originally intended to provide the basis for an environmental impact statement to assist ERDA in making decisions with respect to possible LWR fuel reprocessing and recycling programs. Since the Administration has recently made a decision to indefinitely defer reprocessing, this environmental impact statement is no longer needed. Nevertheless, this document is issued as a report to assist the public in its consideration of nuclear power issues. The statement compares the various alternatives for the LWR fuel cycle. Costs and environmental effects are compared. Safeguards for plutonium from sabotage and theft are analyzed

Characteristic test technology for PWR fuel and its components

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-01-01

Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

Reduced fuel consumption for fork-lift trucks with hydrostatic transmission

Energy Technology Data Exchange (ETDEWEB)

Abels, T

1983-05-01

Cost calculations for a 3,5-t diesel fork lifter done on the basis of VDI 2695 shows, that fuel costs account only for a small part of the operating costs despite the price increase for diesel fuel. Fork lifters with disk-cam controlled primary/secondary adjusted hydrostatic transmission used less fuel than was indicated in the VDI-guideline. Fuel consumption could further be reduced by an optimized hydraulic adjustment together with a precisely harmonized engine speed adjustment. Annual cost savings are considerable.

Survey of nuclear fuel cycle economics: 1970--1985

International Nuclear Information System (INIS)

Prince, B.E.; Peerenboom, J.P.; Delene, J.G.

1977-03-01

This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

Science.gov (United States)

Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

2018-01-01

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

Westinghouse introduces new fuel for PWRs and BWRs

Energy Technology Data Exchange (ETDEWEB)

Orr, W L; McClintock, D C

1985-09-01

In response to utility demands for improved fuel performance, reduced fuel cycle costs, and enhanced operating margins, Westinghouse recently introduced advanced fuel assembly designs for both types of LWR - Vantage 5 for PWRs, and Quad+ for BWRs.

EURLIB-LWR-45/16 and - 15/5. Two board group libraries for LWR-shielding problems

Energy Technology Data Exchange (ETDEWEB)

Herrnberger, V

1982-04-01

Specifications of the broad group cross section libraries EURLIB-LWR-45/16 and -15/5 are given. They are based on EURLIB-III data and produced for LWR shielding problems. The elements considered are H, C{sub 12}, O, Na, Al, Si, Ca, Cr, Mn, Fe, Ni, Zr, U{sub 235}, U{sub 238}. The cross section libraries are available upon request from EIR, RSIC, NEA-CPL and IAEA-NDS. (author) Refs, figs, tabs

Effects of salvage logging and pile-and-burn on fuel loading, potential fire behaviour, fuel consumption and emissions

Science.gov (United States)

Morris C. Johnson; Jessica E. Halofsky; David L. Peterson

2013-01-01

We used a combination of field measurements and simulation modelling to quantify the effects of salvage logging, and a combination of salvage logging and pile-and-burn fuel surface fuel treatment (treatment combination), on fuel loadings, fire behaviour, fuel consumption and pollutant emissions at three points in time: post-windstorm (before salvage logging), post-...

Fuel design and engineering

International Nuclear Information System (INIS)

Hiemer, H.

1975-01-01

The essential aspects of the design and engineering of fuel assemblies for LWR reactors are outlined, and the major criteria to be met by the materials used are given. The fuel rods must be mechanically designed to withstand many stresses which are shortly dealt with here. (RB) [de

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

Energy Technology Data Exchange (ETDEWEB)

Gilles Youinou; R. Sonat Sen

2013-09-01

The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Investigation of the growth rate for joint fast breeder reactor and light water reactor operation

International Nuclear Information System (INIS)

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

1977-07-01

An investigation of fuel consumption and breeding characteristics of FBR-LWR joint operation is presented. The FBR operates in a closed cycle with joint-reprocessing of core and blanket material. The LWR-portion that runs on FBR plutonium operates in an open cycle. The growth rate of the system is defined based upon the fact that the discharge from the system will make up a fraction of an identical system; the system growth rate is found to have an almost linear dependence on the fraction of the LWR fed by plutonium from the FBR. The LWR growth rate, which is negative, is a constant and represents the fraction of the fuel burnt in the LWR-fraction that runs on FBR plutonium per year

Fuel elements for LWR power plants

International Nuclear Information System (INIS)

Roepenack, H.

1977-01-01

About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de

Fuel self-sufficient and low proliferation risk multi-recycling of spent fuel

International Nuclear Information System (INIS)

Cho, N. Z.; Hong, S. G.; Kim, T. H.; Greenspan, E.; Kastenberg, W. E.

1998-01-01

A preliminary feasibility study has been performed in search of promising nuclear energy systems which could make efficient use of the spent fuel from LWRs and be proliferation resistant. The energy considered consist of a dry process and a fuel-self-sufficient reactor which are synergistic. D 2 O, H 2 O and Pb (or Pb-Bi) are considered for the coolant. The most promising identified consists of Pb-cooled reactors with either an AIROX or an IFR-like reprocessing. H 2 O- (possibly mixed with D 2 O) cooled reactors can be designed to be fuel-self-sufficient and multi-recycle LWR spent fuel, provided they are accelerator driven. Moderator-free, D 2 O-cooled critical reactors can multi-recycle Th- 233 U fuel using IFR-type reprocessing; they are significantly more attractive than their thermal counterparts. H 2 O- (possibly mixed with D 2 O) cooled, accelerator-driven reactors appear attractive for converting Th into denatured 233 U using LWR spent fuel and the IFR process. The CANDU reactor technology appears highly synergistic with accelerator-driven systems. (author). 25 refs., 3 tabs., 6 figs

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

Field Measurement and Calibration of HDM-4 Fuel Consumption Model on Interstate Highway in Florida

Directory of Open Access Journals (Sweden)

Xin Jiao

2015-03-01

Full Text Available Fuel consumptions are measured by operating passenger car and tractor-trailer on two interstate roadway sites in Florida. Each site contains flexible pavement and rigid pavement with similar pavement, traffic and environmental condition. Field test reveals that the average fuel consumption differences between vehicle operating on flexible pavement and rigid pavement at given test condition are 4.04% for tractor-trailer and 2.50% for passenger car, with a fuel saving on rigid pavement. The fuel consumption differences are found statistically significant at 95% confidence level for both vehicle types. Test data are then used to calibrate the Highway Development and Management IV (HDM-4 fuel consumption model and model coefficients are obtained for three sets of observations. Field measurement and prediction by calibrated model shows generally good agreement. Nevertheless, verification and adjustment with more experiment or data sources would be expected in future studies.

Linear accelerator fuel enricher regenerator (LAFER) and fission product transmutor (APEX)

International Nuclear Information System (INIS)

Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

1979-01-01

In addition to safety, two other major problems face the nuclear industry today; first is the long-term supply of fissle material and second is the disposal of long-lived fission product waste. The higher energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator interact with a molten lead target to produce spallation and evaporation neutrons. The neutrons are absorbed in a surrounding blanket of light water power reactor (LWR) fuel elements to produce fissile Pu-239 or U-233 fuel from natural fertile U-238 or Th-232 contained in the elements. The fissile enriched fuel element is used in the LWR power reactor until its reactivity is reduced after which the element is regenerated in the linear accelerator target/blanket assembly and then the element is once again burned (fissioned) in the power LWR. In this manner the natural uranium fuel resource can supply an expanding nuclear power reactor economy without the need for fuel reprocessing, thus satisfying the US policy of non-proliferation. In addition, the quantity of spent fuel elements for long-term disposal is reduced in proportion to the number of fuel regeneration cycles through the accelerator. The limiting factor for in-situ regeneration is the burnup damage to the fuel cladding material. A 300 ma-1.5 GeV (450 MW) proton linear accelerator can produce approximately one ton of fissile (Pu-239) material annually which is enough to supply fuel to three 1000 MW(e) LWR power reactors. With two cycles of enriching and regenerating, the nuclear fuel natural resource can be stretched by a factor of 3.6 compared to present fuel cycle practice without the need for reprocessing. Furthermore, the need for isotopic enrichment facilities is drastically reduced

Magnetic scanning of LWR fuel assemblies

International Nuclear Information System (INIS)

Fiarman, S.; Moodenbaugh, A.

1980-01-01

Nondestructive assay (NDA) techniques are available both for fresh and spent fuel, but generally are too time consuming and do not uniquely identify an assembly. A new method is reported to obtain a signature from a magnetic scan of each assembly. This scan is an NDA technique that detects magnetic inclusions. It is potentially fast (5 min/assembly), and may provide a unique signature from the magnetic properties of each fuel assembly

Impact of idling on fuel consumption and exhaust emissions and available idle-reduction technologies for diesel vehicles – A review

International Nuclear Information System (INIS)

Rahman, S.M. Ashrafur; Masjuki, H.H.; Kalam, M.A.; Abedin, M.J.; Sanjid, A.; Sajjad, H.

2013-01-01

Highlights: • In this paper we reviewed the impact of diesel vehicles idling on fuel consumption and exhaust emission. • Fuel consumption and emissions during idling are very high compared to driving cycle. • The effects of various operating on fuel consumption and exhaust emission were discussed. • Available idle-reduction technologies impact on idling fuel consumption and emissions were discussed. • Idling reduction technologies reduce fuel consumption and emissions significantly. - Abstract: In order to maintain cab comfort truck drivers have to idle their engine to obtain the required power for accessories, such as the air conditioner, heater, television, refrigerator, and lights. This idling of the engine has a major impact on its fuel consumption and exhaust emission. Idling emissions can be as high as 86.4 g/h, 16,500 g/h, 5130 g/h, 4 g/h, and 375 g/h for HC, CO 2 , CO, PM, and NOx, respectively. Idling fuel consumption rate can be as high as 1.85 gal/h. The accessory loading, truck model, fuel-injection system, ambient temperature, idling speed, etc., also affect significantly the emission levels and fuel consumption rate. An increase in accessory loading and ambient temperature increases the emissions and fuel consumption. During idling, electronic fuel-injection systems reduce HC, PM, and CO emission, but increase NOx emissions compared with a mechanical fuel-injection system. An increase of idling speed increases fuel consumption rate. There are many systems available on the market to reduce engine idling and improve air quality and fuel consumption rate, such as an auxiliary power unit (APU), truck stop electrification, thermal storage systems, fuel cells, and direct fire heaters. A direct fire heater reduces fuel consumption by 94–96% and an APU reduces consumption by 60–87%. Furthermore, these technologies increase air quality significantly by reducing idling emissions, which is the reason why they are considered as key alternatives to

75 FR 66008 - Fossil Fuel-Generated Energy Consumption Reduction for New Federal Buildings and Major...

Science.gov (United States)

2010-10-27

... Fossil Fuel-Generated Energy Consumption Reduction for New Federal Buildings and Major Renovations of Federal Buildings; Correction AGENCY: Office of Energy Efficiency and Renewable Energy, Department of... the fossil fuel- generated energy consumption [[Page 66009

Fluidized-bed calcination of LWR fuel-reprocessing HLLW: requirements and potential for off-gas cleanup

International Nuclear Information System (INIS)

Schindler, R.E.

1979-01-01

Fluidized-bed solidification (calcination) was developed on a pilot scale for a variety of simulated LWR high-level liquid-waste (HLLW) and blended high-level and intermediate-level liquid-waste (ILLW) compositions. It has also been demonstrated with ICPP fuel-reprocessing waste since 1963 in the Waste Calcining Facility (WCF) at gross feed rates of 5 to 12 m 3 /day. A fluidized-bed calciner produces a relatively large volume of off-gas. A calciner solidifying 6 m 3 /day of liquid waste would generate about 13 standard m 3 /min of off-gas containing 10 to 20 g of entrained solids per standard m 3 of off-gas. Use of an off-gas system similar to that of the WCF could provide an overall process decontamination factor for particulates of about 2 x 10 10 . A potential advantage of fluidized-bed calcination over other solidification methods is the ability to control ruthenium volatilization from the calciner at less than 0.01% by calcining at 500 0 C or above. Use of an off-gas system similar to that of the WCF would provide an overall process decontamination factor for volatile ruthenium of greater than 1.6 x 10 7

An Efficient Two-Objective Hybrid Local Search Algorithm for Solving the Fuel Consumption Vehicle Routing Problem

OpenAIRE

Rao, Weizhen; Liu, Feng; Wang, Shengbin

2016-01-01

The classical model of vehicle routing problem (VRP) generally minimizes either the total vehicle travelling distance or the total number of dispatched vehicles. Due to the increased importance of environmental sustainability, one variant of VRPs that minimizes the total vehicle fuel consumption has gained much attention. The resulting fuel consumption VRP (FCVRP) becomes increasingly important yet difficult. We present a mixed integer programming model for the FCVRP, and fuel consumption is ...

The Fuel Performance Analysis of LWR Fuel containing High Thermal Conductivity Reinforcements

International Nuclear Information System (INIS)

Kim, Seung Su; Ryu, Ho Jin

2015-01-01

The thermal conductivity of fuel affects many performance parameters including the fuel centerline temperature, fission gas release and internal pressure. In addition, enhanced safety margin of fuel might be expected when the thermal conductivity of fuel is improved by the addition of high thermal conductivity reinforcements. Therefore, the effects of thermal conductivity enhancement on the fuel performance of reinforced UO2 fuel with high thermal conductivity compounds should be analyzed. In this study, we analyzed the fuel performance of modified UO2 fuel with high thermal conductivity reinforcements by using the FRAPCON-3.5 code. The fissile density and mechanical properties of the modified fuel are considered the same with the standard UO2 fuel. The fuel performance of modified UO2 with high thermal conductivity reinforcements were analyzed by using the FRAPCON-3.5 code. The thermal conductivity enhancement factors of the modified fuels were obtained from the Maxwell model considering the volume fraction of reinforcements

Hydrogen as a renewable and sustainable solution in reducing global fossil fuel consumption

International Nuclear Information System (INIS)

Midilli, Adnan; Dincer, Ibrahim

2008-01-01

In this paper, hydrogen is considered as a renewable and sustainable solution for reducing global fossil fuel consumption and combating global warming and studied exergetically through a parametric performance analysis. The environmental impact results are then compared with the ones obtained for fossil fuels. In this regard, some exergetic expressions are derived depending primarily upon the exergetic utilization ratios of fossil fuels and hydrogen: the fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency, fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator. These relations incorporate predicted exergetic utilization ratios for hydrogen energy from non-fossil fuel resources such as water, etc., and are used to investigate whether or not exergetic utilization of hydrogen can significantly reduce the fossil fuel based global irreversibility coefficient (ranging from 1 to +∞) indicating the fossil fuel consumption and contribute to increase the hydrogen based global exergetic indicator (ranging from 0 to 1) indicating the hydrogen utilization at a certain ratio of fossil fuel utilization. In order to verify all these exergetic expressions, the actual fossil fuel consumption and production data are taken from the literature. Due to the unavailability of appropriate hydrogen data for analysis, it is assumed that the utilization ratios of hydrogen are ranged between 0 and 1. For the verification of these parameters, the variations of fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator as the functions of fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency and exergetic utilization of hydrogen from non-fossil fuels are analyzed and discussed in detail. Consequently, if exergetic utilization ratio of hydrogen from non-fossil fuel sources at a certain exergetic utilization ratio of fossil fuels increases

Wood fuel consumption and mortality rates in Sub-Saharan Africa: Evidence from a dynamic panel study.

Science.gov (United States)

Sulaiman, Chindo; Abdul-Rahim, A S; Chin, Lee; Mohd-Shahwahid, H O

2017-06-01

This study examined the impact of wood fuel consumption on health outcomes, specifically under-five and adult mortality in Sub-Saharan Africa, where wood usage for cooking and heating is on the increase. Generalized method of moment (GMM) estimators were used to estimate the impact of wood fuel consumption on under-five and adult mortality (and also male and female mortality) in the region. The findings revealed that wood fuel consumption had significant positive impact on under-five and adult mortality. It suggests that over the studied period, an increase in wood fuel consumption has increased the mortality of under-five and adult. Importantly, it indicated that the magnitude of the effect of wood fuel consumption was more on the under-five than the adults. Similarly, assessing the effect on a gender basis, it was revealed that the effect was more on female than male adults. This finding suggests that the resultant mortality from wood smoke related infections is more on under-five children than adults, and also are more on female adults than male adults. We, therefore, recommended that an alternative affordable, clean energy source for cooking and heating should be provided to reduce the wood fuel consumption. Copyright © 2017 Elsevier Ltd. All rights reserved.

How hybrid-electric vehicles are different from conventional vehicles: the effect of weight and power on fuel consumption

International Nuclear Information System (INIS)

Reynolds, C; Kandlikar, M

2007-01-01

An increasingly diverse set of hybrid-electric vehicles (HEVs) is now available in North America. The recent generation of HEVs have higher fuel consumption, are heavier, and are significantly more powerful than the first generation of HEVs. We compare HEVs for sale in the United States in 2007 to equivalent conventional vehicles and determine how vehicle weight and system power affects fuel consumption within each vehicle set. We find that heavier and more powerful hybrid-electric vehicles are eroding the fuel consumption benefit of this technology. Nonetheless, the weight penalty for fuel consumption in HEVs is significantly lower than in equivalent conventional internal combustion engine vehicles (ICEVs). A 100 kg change in vehicle weight increases fuel consumption by 0.7 l/100 km in ICEVs compared with 0.4 l/100 km in HEVs. When the HEVs are compared with their ICEV counterparts in an equivalence model that differentiates between cars and sports-utility vehicles, the average fuel consumption benefit was 2.7 l/100 km. This analysis further reveals that a HEV which is 100 kg heavier than an identical ICEV would have a fuel consumption penalty of 0.15 l/100 km. Likewise, an increase in the HEV's power by 10 kW results in a fuel consumption penalty of 0.27 l/100 km

Corrosion Tests of LWR Fuels - Nuclide Release

International Nuclear Information System (INIS)

P.A. Finn; Y. Tsai; J.C. Cunnane

2001-01-01

Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The 99 Tc, 129 I, 137 Cs, 97 Mo, and 90 Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the 99 Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup

Fabrication of Multi-Layerd SiC Composite Tube for LWR Applications

Energy Technology Data Exchange (ETDEWEB)

Kim, Daejong; Jung, Choonghwan; Kim, Weonju; Park, Jiyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Jongmin [Chungnam National Univ., Daejeon (Korea, Republic of)

2013-05-15

In this study, the chemical vapor deposition (CVD) and chemical vapor infiltration (CVI) methods were employed for the fabrication of the composite tubes. SiC ceramics and SiC-based composites have recently been studied for LWR fuel cladding applications because of good mechanical/physical properties, neutron irradiation resistance and excellent compatibility with coolant under severe accident. A multi-layered SiC composite tube as the nuclear fuel cladding is composed of the monolith SiC inner layer, SiC/SiC composite intermediate layer, and monolith SiC outer layer. Since all constituents should be highly pure, stoichiometric to achieve the good properties, it has been considered that the chemical process is a well-suited technique for the fabrication of the SiC phases.

Spent LWR fuel-storage costs

International Nuclear Information System (INIS)

Clark, H.J.

1981-01-01

Expanded use of existing storage basins is clearly the most economic solution to the spent fuel storage problem. The use of high-density racks followed by fuel disassembly and rod storage is an order of magnitude cheaper than building new facilities adjacent to the reactor. The choice of a new storage facility is not as obvious; however, if the timing of expenditures and risk allowance are to be considered, then modular concepts such as silos, drywells, and storage casks may cost less than water basins and air-cooled vaults. A comparison of the costs of the various storage techniques without allowances for timing or risk is shown. The impact of allowances for discounting and early resumption of reprocessing is also shown. Economics is not the only issue to be considered in selecting a storage facility. The licensing, environmental impact, timing, and social responses must also be considered. Each utility must assess all of these issues for their particular reactors before the best storage solution can be selected

Method for pre-processing LWR spent fuel

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Ebihara, Hikoe.

1986-01-01

Purpose: To facilitate the decladding of spent fuel, cladding tube processing, and waste gas recovery, and to enable the efficient execution of main re-processing process thereafter. Constitution: Spent fuel assemblies are sent to a cutting process where they are cut into chips of easy-to-process size. The chips, in a thermal decladding process, undergo a thermal cycle processing in air with the processing temperatures increased and decreased within the range of from 700 deg C to 1200 deg C, oxidizing zircaloy comprising the cladding tubes into zirconia. The oxidized cladding tubes have a number of fine cracks and become very brittle and easy to loosen off from fuel pellets when even a slight mechanical force is applied thereto, thus changing into a form of powder. Processed products are then separated into zirconia sand and fuel pellets by a gravitational selection method or by a sifting method, the zirconia sand being sent to a waste processing process and the fuel pellets to a melting-refining process. (Yoshino, Y.)

Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

International Nuclear Information System (INIS)

Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

1997-03-01

This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide

Cracking and relocation of UO2 fuel during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Dagbjartsson, S.J.

1981-01-01

Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

International Nuclear Information System (INIS)

Rasmussen, D.E.; Schaus, P.S.

1980-01-01

The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

A method of neptunium recovery into the product stream of the Purex 1st codecontamination step for LWR fuel reprocessing

International Nuclear Information System (INIS)

Tsuboya, Takao; Nemoto, Shinichi; Hoshino, Tadaya; Segawa, Takeshi

1973-01-01

An improved nitrous acid method was applied for recovering neptunium in spent fuel. Counter-current solvent extraction has been performed to find out its recovery conditions. The nitrous acid in the form of sodium salt solution was fed to the 1st stage of extraction section, and hydrazine nitrate was fed to some stages near feed point. Flow rate and the concentration of additives were altered for finding out optimum condition. Laboratory scale mixer-settlers having 6 ml of mixing volume and 17 ml of settling volume for each stage were used. The nitrous acid method was improved so that the reduction reaction in scrub section can be eliminated by the decomposition of the nitrous acid using a reagent such as sulfamic acid, urea, or hydrazine. In operation, the feed rate of the nitrous acid was about 3 mM/hr, and about 61% of neptunium charged was discharged in the product stream of Purex-1st codecontamination step designed for the LWR fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation. The calculated value of Δx/x for extraction section agreed with the experimental value, where Δx is the quantity of oxidation, and x is the inventory for neptunium in each stage. In conclusion, the improved nitrous acid method is effective for the neptunium discharge in product stream, and the difference of neptunium extraction between estimate and experiment is caused by some of reduction reaction in scrub section. (Iwakiri, K.)

Transmutation of LWR waste actinides in thermal reactors

International Nuclear Information System (INIS)

Gorrell, T.C.

1979-01-01

Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-05-01

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

Evaluation and optimization of LWR fuel cycles

International Nuclear Information System (INIS)

Akbas, T.; Zabunoglu, O.; Tombakoglu, M.

2001-01-01

There are several options in the back-end of the nuclear fuel cycle. Discharge burn-up, length of interim storage period, choice of direct disposal or recycling and method of reprocessing in case of recycling affect the options and determine/define the fuel cycle scenarios. These options have been evaluated in viewpoint of some tangible (fuel cycle cost, natural uranium requirement, decay heat of high level waste, radiological ingestion and inhalation hazards) and intangible factors (technological feasibility, nonproliferation aspect, etc.). Neutronic parameters are calculated using versatile fuel depletion code ORIGEN2.1. A program is developed for calculation of cost related parameters. Analytical hierarchy process is used to transform the intangible factors into the tangible ones. Then all these tangible and intangible factors are incorporated into a form that is suitable for goal programming, which is a linear optimization technique and used to determine the optimal option among alternatives. According to the specified objective function and constraints, the optimal fuel cycle scenario is determined using GPSYS (a linear programming software) as a goal programming tool. In addition, a sensitivity analysis is performed for some selected important parameters

A classification scheme for LWR fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Moore, R.S.; Williamson, D.A.; Notz, K.J.

1988-11-01

With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

A classification scheme for LWR fuel assemblies

International Nuclear Information System (INIS)