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Sample records for integral fuel burnable

  1. Benchmark solution of contemporary PWR integral fuel burnable absorbers

    International Nuclear Information System (INIS)

    Stucker, D.L.; Hone, M.J.; Holland, R.A.

    1993-01-01

    This paper presents a closely controlled benchmark solution of the two major contemporary pressurized water reactor integral burnable absorber designs: zirconium diboride (ZrB 2 ) and gadolinia (Gd 2 O 3 ). The comparison is accomplished using self-generating equilibrium cycles with equal energy, equal discharge burnup, and equal safety constraints. The reference plant for this evaluation is a 3411-MW(thermal) Westinghouse four-loop nuclear steam supply system operating with an inlet temperature of 285.9 degrees C, a core coolant mass now rate of 16877.3 kg/s, and coolant pressure of 15.5 MPa. The reactor consists of 193 VANTAGE 5H fuel assemblies that are discharged at a region average burnup of 48.4 GWd/tonne U. Each fuel assembly contains a natural uranium axial blanket 15.24 cm long at the top and the bottom of the fuel rod. The burnable absorber rods are symmetrically radially dispersed within the fuel assembly such that intrabundle power peaking is minimized. The burnable absorber material for both ZrB 2 and Gd 2 O 3 is axially zoned to the central 304.8 cm of the absorber-bearing fuel rods. The fuel management was constrained such that the thermal and safety limitations of F δH q -5 /degrees C were simultaneously achieved. The maximum long-term operating soluble boron concentration was also limited to 446 effective full-power days (EFPDs) including 14 EFPDs of power coastdown were assumed

  2. Burnable absorber coated nuclear fuel

    International Nuclear Information System (INIS)

    Chubb, W.; Radford, K.C.; Parks, B.H.

    1984-01-01

    A nuclear fuel body which is at least partially covered by a burnable neutron absorber layer is provided with a hydrophobic overcoat generally covering the burnable absorber layer and bonded directly to it. In a method for providing a UO 2 fuel pellet with a zirconium diboride burnable poison layer, the fuel body is provided with an intermediate niobium layer. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. Burnable poison fuel element and its fabrication

    International Nuclear Information System (INIS)

    Zukeran, Atsushi; Inoue, Kotaro; Aizawa, Hiroko.

    1985-01-01

    Purpose: To enable to optionally vary the excess reactivity and fuel reactivity. Method: Burnable poisons with a large neutron absorption cross section are contained in fuel material, by which the excess reactivity at the initial stage in the reactor is suppressed by the burnable poisons and the excess reactivity is released due to the reduction in the atomic number density of the burnable poisons accompanying the burning. The burnable poison comprises spherical or rod-like body made of a single material or spherical or rod-like member made of a plurality kind of materials laminated in a layer. These spheres or rods are dispersed in the fuel material. By adequately selecting the shape, combination and the arrangement of the burnable poisons, the axial power distribution of the fuel rods are flattened. (Moriyama, K.)

  5. Fuel assembly and burnable poison rod

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1993-01-01

    In a fuel assembly having burnable poison rods arranged therein, the burnable poison comprises an elongate small outer tube and an inner tube coaxially disposed within the outer tube. Upper and lower end tubes each sealed at one end are connected to both of the upper and lower ends in the inner and the outer tubes respectively. A coolant inlet hole is disposed to the lower end tube, while a coolant leakage hole is disposed to the upper end tube. Burnable poison members are filled in an annular space. Further, the burnable poison-filling region is disposed excepting portions for 1/20 - 1/12 of the effective fuel length at each of the upper and the lower ends of the fuel rod. Then, the concentration of the burnable poisons in a region above a boundary defined at a position 1/3 - 1/2, from beneath, of the effective fuel length is made smaller than that in the lower region. This enables to suppress excess reactions of fuels to reduce the mass of the burnable neutron. Excellent reactivity control performance at the initial stage of the burning can be attained. (T.M.)

  6. Burnable absorber-integrated Guide Thimble (BigT) - 1. Design concepts and neutronic characterization on the fuel assembly benchmarks

    International Nuclear Information System (INIS)

    Yahya, Mohd-Syukri; Yu, Hwanyeal; Kim, Yonghee

    2016-01-01

    This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named 'Burnable absorber-integrated Guide Thimble' (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal-hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17x17 and 16x16 fuel assembly lattices. For the 17x17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16x16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library. (author)

  7. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Pitts, M.L.

    2000-01-01

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

  8. Burnable poison option for DUPIC fuel

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Cupta, H. P.

    1996-08-01

    The mechanisms of positive coolant void reactivity of CANDU natural uranium and DUPIC fuel have been studied. The design study of DUPIC fuel was performed using the burnable poison material in the center pin to reduce the coolant void reactivity. The amount of burnable poison was determined such that the prompt inverse period of DUPIC fuel upon full coolant voiding is the same as that of natural uranium fuel at equilibrium burnup. A parametric study on various burnable poisons has shown that natural dysprosium has more merit over other materials because it uniformly controls the void reactivity throughout the burnup with reasonable amount of poison. Additional studies on the option of using scattering or absorber material in the center pin position and the option using variable fuel density were performed. In any case of option using variable fuel density were performed. In any case of options to reduce the void reactivity, it was found that either the discharge burnup and/or the relative linear pin power are sacrificed. A preliminary study was performed for the evaluation of reference DUPIC fuel performance especially represented by Stress Corrosion Cracking(SCC) parameters which is mainly influenced by the refueling operations. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increment of the reference DUPIC fuel are below the SCC thresholds of CANDU natural uranium fuel. For a 4-bundle shift refueling scheme, the envelopes of element ramped power and power increment upon refueling are 8% and 44% higher than those of a 2-bundle shift refueling scheme on the average, respectively, but still have margins to the failure thresholds of natural uranium fuel. 23 tabs., 25 figs., 20 refs. (Author)

  9. Group constants calculation for fuel assemblies containing burnable absorbers; Prorachun grupnih konstanti gorivnih elemenata koji sadrzhe sagorive apsorbere

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B [Institut Rudjer Boskovic, Zagreb (Yugoslavia); Pevec, D [Elektrotehnicki Fakultet, Zagreb Univ. (Yugoslavia); Urli, N; Shmuc, T [Institut Rudjer Boskovic, Zagreb (Yugoslavia)

    1988-07-01

    The upgrading of the computer code package PSU-LEOPARD/MCRAC is described. The upgraded package enables modelling of fuel assemblies containing burnable absorbers in the form of borosilicate glass rodlets, or, integral fuel burnable absorbers. The package is tested using the NPP Krsko core data. (author)

  10. Production method of burnable poison incorporated fuel pellet by coating

    International Nuclear Information System (INIS)

    Naito, Naoyoshi.

    1993-01-01

    A cylindrical member is formed with an organic material which is melted, decomposed or evaporated by heating. Such organic materials include polyethylene and polyvinyl alcohol, for example. A predetermined amount of burnable poisons are homogeneously incorporated in the cylindrical member by a means, such as melting before fabricating it into a cylindrical shape. UO 2 fuel pellets are inserted to the cylindrical member and heated, to scatter only the organic materials, so that non-volatile burnable poisons are homogeneously left on the surface of the pellets. It is preferred that the cylindrical member having pellets inserted therein is inserted to a cladding tube and applied with a heat treatment. With such procedures, a UO 2 pellet is coated with burnable poisons by a convenient and compact device. In addition, grinding step after the coating is unnecessary. (I.N.)

  11. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  12. Burnable poison rod for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Funk, C.E.; Oneufer, A.S.

    1984-01-01

    A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

  13. The burnable poisons utilization for fissile enriched CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

    1996-12-31

    Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.

  14. Impact of burnable absorber Gd on nuclide composition for VVER-440 fuel (Gd-2)

    International Nuclear Information System (INIS)

    Zajac, R.; Chrapciak, V.

    2010-01-01

    The latest version of Russian fuel VVER-440 includes burnable absorber in 6 pins. In this article is impact of burnable absorber on nuclide composition and criticality analyzed. In part 1 was analyzed whole burnup interval 0-50 MWd/kgU. In present part 2 are detailed analysis only for first cycle (burnup 0-10 MWd/kgU). (Authors)

  15. A model for fuel shuffling and burnable absorbers optimization in low leakage PWRs

    International Nuclear Information System (INIS)

    Zavaljevski, N.

    1990-01-01

    A nonlinear model for the simultaneous optimization of fuel shuffling and burnable absorbers in PWRs is formulated using the depletion perturbation theory. The sensitivity coefficients are defined in a new way, using a macroscopic burnup model coupled with the explicit burnable absorbers depletion equation. Since first-order perturbation theory is limited to small changes in burnable absorber concentration, the associated control variable is continuous, with a constraint on maximal increment. Fuel shuffling is described by Boolean variables. Thus a special case of a mixed-integer quadratic programming problem is obtained, since the interaction of fuel and absorber optimization is considered. (author)

  16. Neutronic analysis of a fuel element with variations in fuel enrichment and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Martins, Felipe; Velasquez, Carlos E.; Cardoso, Fabiano; Fortini, Angela; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In this work, the goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers ({sup 155}Gd and {sup 157}Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn. (author)

  17. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  18. Calculational modeling of fuel assemblies of WWER-1000 type with the use of burnable absorber Gadolinum; comparative analysis

    International Nuclear Information System (INIS)

    Yeremenko, M.L.; Kovbasenko, Yu.P.; Loetsch, T.

    2001-01-01

    In connection with the beginning of the use of fuel assemblies with burnable absorbers by integration of Gadolinum into the nuclear fuel at Ukrainian NPP the task of testing the code systems and the pertinent neutron cross section libraries for the new fuel arose. Taking into account the long term experience of German experts with calculations and evaluation of nuclear fuel containing Gadolinum it was decided to carry out a series of test calculations for fuel assembly lattices of PWR, WWER-440 and WWER-1000 types using the NESSEL/PYTHIA and CASMO/SIMULATE code systems (Authors)

  19. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1985-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  20. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Nagaoka, Yoshiharu; Oyamada, Rokuro [Japan Atomic Energy Research Institute, Oarai-machi Ibaraki-ken (Japan); Matos, J E; Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  1. Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

    International Nuclear Information System (INIS)

    Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

    2009-01-01

    The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.

  2. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1992-01-01

    This patent describes a burnable absorber coated nuclear fuel. It comprises a nuclear fuel substrate containing a fissionable material; and an outer burnable absorber coating applied on an outer surface of the substrate; the outer absorber coating being composed of an inner layer of a boron-bearing material except for erbium boride and an outer layer of an erbium material

  3. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  4. The manufacture process and properties of (U, Gd)O2 burnable poisonous fuel pellets

    International Nuclear Information System (INIS)

    Yi Wei; Tang Yueming; Dai Shengping; Yang Youqing; Zuo Guoping; Wu Shihong; Gu Xiaofei; Gu Mingfei

    2006-03-01

    The main properties of important raw powder materials used in the (U, Gd)O 2 burnable poisonous fuel pellets production line of NPIC are presented. The powders included UO 2 , Gd 2 O 3 , (U, Gd) 3 O 8 and necessary additives, such as ammonium oxalate and zinc stearate. And the main properties of (U, Gd)O 2 burnable poisonous fuel pellets and the manufacture processes, such as ball-milling blending, granulation, pressing, sintering and grinding are also described. Moreover, the main effect of the process parameters controlled in the manufacture process have been discussed. (authors)

  5. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1984-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs

  6. Fuel with advanced burnable absorbers design for the IRIS reactor core: Combined Erbia and IFBA

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, Fausto [Westinghouse Electric Company LLC, Science and Technology Department, Pittsburgh, PA 15235 (United States)], E-mail: FranceF@westinghouse.com; Petrovic, Bojan [Georgia Institute of Technology, Nuclear and Radiological Engineering, G.W. Woodruff School, Atlanta, GA 30332-0405 (United States)

    2009-08-15

    IRIS is an advanced medium-size (1000 MW) PWR with integral primary system targeting deployment already around 2015-2017. Consistent with its aggressive development and deployment schedule, the 'first IRIS' core design assumes current, licensed fuel technology, i.e., UO{sub 2} fuel with less than 5% {sup 235}U enrichment. The core consists of 89 fuel assemblies employing the 17x17 Westinghouse Robust Fuel Assembly (RFA) design and Standard Fuel dimensions. The adopted design enables to meet all the objectives of the first IRIS core, including over 3-year cycle length with low soluble boron concentration, within the envelope of licensed, readily available fuel technology. Alternative fuel designs are investigated for the subsequent waves of IRIS reactors in pursuit of further improving the fuel utilization and/or extending the cycle length. In particular, an increase in the lattice pitch from the current 0.496 in. for the Standard Fuel to 0.523 in. is among the objectives of this study. The larger fuel pitch and increased moderator-to-fuel volume ratio that it entails fosters better neutron thermalization in an altogether under-moderated lattice thereby offering the potential for considerable increase of fuel utilization and cycle length, up to 5% in the two-batch fuel management scheme considered for IRIS. However, the improved moderation also favors higher values of the Moderator Temperature Coefficient, MTC, which must be properly counteracted to avoid undesired repercussions on the plant safety parameters or controllability during transient operations. This paper investigates counterbalancing the increase in the MTC caused by the enhanced moderation lattice by adopting a suitable choice of fuel burnable absorber (BA). In particular, a fuel design combining erbia, which benefits MTC due to its resonant behavior but leads to residual reactivity penalty, and IFBA, which maximizes cycle length, is pursued. In the proposed approach, IFBA provides the bulk

  7. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1991-01-01

    This patent describes a nuclear reactor core. It comprises a first group of fuel rods containing fissionable material and being free of burnable absorber material; and a second group of fuel rods containing fissionable material and first and second burnable absorber material; the first burnable absorber material being a boron-bearing material which does not contain erbium and the second burnable absorber material being an erbium material; the first and second burnable absorber materials being in the form of an outer coating on the fissionable material, the outer coating being composed of an inner layer of one of the boron-bearing material which does not contain erbium and the erbium material and an outer layer of the other of the boron-bearing material which does not contain erbium and the erbium material

  8. Reloading optimization of pressurized water reactor core with burnable absorber fuel

    International Nuclear Information System (INIS)

    Shi Xiuan; Liu Zhihong; Hu Yongming

    2008-01-01

    The reloading optimization problem of PWR with burnable absorber fuel is very difficult, and common optimization algorithms are inefficient and have bad global performance for it. Characteristic statistic algorithm (CSA) is very fit for the problem. In the past, the reloading optimization using CSA has shortcomings of separating the fuel assemblies' loading pattern (LP) optimization from burnable absorber's placement (BP) optimization. In this study, LP and BP were optimized simultaneously using CSA coupled with CYCLE2D, which is a core analysis code. The corresponding reloading coupling optimization software, CSALPBP, was developed. The 10th cycle reloading design of Daya Bay Nuclear Power Plant was optimized using CSALPBP. The results show that CSALPBP has high efficiency and excellent global performance. (authors)

  9. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

    Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

  10. Analysis of burnable poison in Ford Nuclear Reactor fuel to extend fuel lifetime. Final report, August 1, 1994--September 29, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Burn, R.R.; Lee, J.C.

    1996-12-01

    The objective of the project was to establish the feasibility of extending the lifetime of fuel elements for the Ford Nuclear Reactor (FNR) by replacing current aluminide fuel with silicide fuel comprising a heavier uranium loading but with the same fissile enrichment of 19.5 wt% {sup 235}U. The project has focused on fuel designs where burnable absorbers, in the form of B{sub 4}C, are admixed with uranium silicide in fuel plates so that increases in the control reactivity requirements and peak power density, due to the heavier fuel loading, may be minimized. The authors have developed equilibrium cycle models simulating current full-size aluminide core configurations with 43 {approximately} 45 fuel elements. Adequacy of the overall equilibrium cycle approach has been verified through comparison with recent FNR experience in spent fuel discharge rates and simulation of reactor physics characteristics for two representative cycles. Fuel cycle studies have been performed to compare equilibrium cycle characteristics of silicide fuel designs, including burnable absorbers, with current aluminide fuel. These equilibrium cycle studies have established the feasibility of doubling the fuel element lifetime, with minimal perturbations to the control reactivity requirements and peak power density, by judicious additions of burnable absorbers to silicide fuel. Further study will be required to investigate a more practical silicide fuel design, which incorporates burnable absorbers in side plates of each fuel element rather than uniformly mixes them in fuel plates.

  11. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

    Purpose: To increase the reactor core lifetime by decreasing the effect of neutron absorption of burnable poison rods by using material with less neutron absorbing effect. Constitution: Stainless steels used so far as the coating material for burnable poison rods have relatively great absorption in the thermal neutral region and are not preferred in view of the neutron economy. Burnable poison rods having fuel can made of zirconium alloy shows absorption the thermal neutron region lower by one digit than that of stainless steels but they shows absorption in the resonance region and the cost is higher. In view of the above, the fuel can of the burnable poison material is made of aluminum or aluminu alloy. This can reduce the neutron absorbing effect by stainless steel fuel can and effectively utilize neutrons that have been wastefully absorbed and consumed in stainless steels. (Takahashi, M.)

  12. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  13. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    Lerner, Ana Maria; Madariaga, Marcelo

    2002-01-01

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  14. Applying burnable poison particles to reduce the reactivity swing in high temperature reactors with batch-wise fuel loading

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Dam, H. van; Hagen, T.H.J.J. van der

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble with a radius of 3 cm containing 9 g of 8% enriched uranium and burnable poison particles (BPP) made of B 4 C highly enriched in 10 B. The radius of the BPP and the number of particles per fuel pebble have been varied to find the flattest reactivity-to-time curve. It was found that for a k∞ of 1.1, a reactivity swing as low as 2% can be obtained when each fuel pebble contains about 1070 BPP with a radius of 75 μm. For coated BPP that consist of a graphite kernel with a radius of 300 μm covered with a B 4 C burnable poison layer, a similar value for the reactivity swing can be obtained. Cylindrical particles seem to perform worse. In general, the modification of the geometry of BPP is an effective means to tailor the reactivity curve of HTRs

  15. Application of boron and gadolinium burnable poison particles in UO2 and PUO2 fuels in HTRs

    International Nuclear Information System (INIS)

    Kloosterman, J.L.

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble (fuel zone with radius of 2.5 cm surrounded with a 0.5 cm thick graphite layer) and burnable poison particles (BPPs) containing B 4 C made of pure 10 B or containing Gd 2 O 3 made of natural Gd. Two types of fuel were considered: UO 2 fuel made of 8% enriched uranium and PuO 2 fuel made of plutonium from LWR spent fuel. The radius of the BPP and the number of particles per fuel pebble were varied to find the flattest reactivity-to-time curve. For the UO 2 fuel, the reactivity swing is lowest (around 2%) for BPPs made of B 4 C with radius of 75 μm. In this case around 1070 BPPs per fuel pebble are needed. For the PuO 2 fuel to get a reactivity swing below 4%, the optimal radius of the BPP is the same, but the number of particles per fuel pebble should be around 1600. The optimal radius of the Gd 2 O 3 particles in the UO 2 fuel is about 10 times that of the B 4 C particles. The reactivity swing is around 3% when each fuel pebble contains only 9 BPPs with radius of 840 μm. The results of the Gd particles illustrate nicely the usage of black burnable poison particles introduced by Van Dam [Ann. Nuclear Energy 27 (2000) 733

  16. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd 2 O 3 ) with a large number of low concentration gad rods (2 w/o Gd 2 O 3 ). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (18+ months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. These increases in the boron concentration would also require the plant to operate at higher lithium (Li) concentrations in the coolant in order to maintain the pH level at the desired value. Operation at the higher Li concentrations is undesirable because of the concerns over the potential impact on the fuel assembly material performance (e.g., crud and corrosion). This paper also reviews the APA (Alpha/Phoenix-P/ANC) nuclear design code system performance for the low concentration gad design. The design system performance for the reload cores that have or are employing this design has been completely satisfactory. The performance and accuracy of the nuclear design methodology is found to be as good for this design as for the reload cores that use exclusively high gad concentrations, or those that use WABA's - the discrete burnable absorber (BA) used prior to its substitution for gadolinium. (authors)

  17. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Science.gov (United States)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  18. Usage of burnable poison on research reactors

    International Nuclear Information System (INIS)

    Villarino, Eduardo Anibal

    2002-01-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  19. Optimizing the use of gadolinium as burnable poison in nuclear fuel: towards a boron free PWR

    International Nuclear Information System (INIS)

    Pieck, D.

    2013-01-01

    Reactivity excess in Nuclear Power Plants is controlled by reactor's active systems: boric acid dilution and control rods. Alternatively, negative reactivity insertion can be made in a passive way using burnable poisons, i.e. neutron absorbers, this is the case of gadolinium (Gd). In the industrial framework of U 235 enrichment increase and boric acid restraint, the goal of this thesis is to optimize the distribution of gadolinium in UO 2 ceramics to obtain a high-performance provision of negative reactivity in Pressurized Water Reactors. In this sense, the work is focus on new gadolinium-rich materials. Thus, U-Gd-O phase diagram was explored in the field of high Gd contents. Two cubic phases were found and characterized: the C1 and C2 phases. With the aim of an industrial application, C1 phase was selected as candidate for Gd addition into UO 2 pellets. The optimal distribution of C1 phase within a nuclear fuel assembly was studied using APOLLO 2.8 neutron transport code. Parametric calculations were performed. These neutronic studies have ends in a successful 'concept of poisoned pellet'. Finally, some prototype pellets following this concept were made in laboratory to proof it feasibility. All the obtained results shows that the proposed concept of a neutro-phage C1-phase coating on UO 2 pellets is a convenient way to reduce reactivity excess within the framework of long irradiation cycles. This concept could be potentially applied in industrial scale. Consequently a patent application process was initiated.(author) [fr

  20. Optimization of burnable poison disposition for in-core fuel assemblies

    International Nuclear Information System (INIS)

    Zhong Wenfa; Luo Rong; Zhou Quan

    1997-09-01

    The optimization of the burnable poison disposition in the initial core loading of the 200 MW nuclear heating reactor (NHR-200), is studied. The mass fraction of the burnable poison is used as the control variable with the objective to minimize the power peaking factor. The flexible tolerance method is used to solve the nonlinear programming optimal problem. The optimization method can be used in reactor physics design, and get a new pattern of initial core which is of reference value. (2 refs., 8 figs., 1 tab.)

  1. Heterogeneous burnable poisons:

    International Nuclear Information System (INIS)

    Leiva, Sergio; Agueda, Horacio; Russo, Diego

    1989-01-01

    The use of materials possessing high neutron absorption cross-section commonly known as 'burnable poisons' have its origin in BWR reactors with the purpose of improving the efficiency of the first fuel load. Later on, it was extended to PWR to compensate of initial reactivity without infringing the requirement of maintaining a negative moderator coefficient. The present tendency is to increase the use of solid burnable poisons to extend the fuel cycle life and discharge burnup. There are two concepts for the burnable poisons utilization: 1) heterogeneously distributions in the form of rods, plates, etc. and 2) homogeneous dispersions of burnable poisons in the fuel. The purpose of this work is to present the results of sinterability studies, performed on Al 2 O 3 -B 4 C and Al 2 O 3 -Gd 2 O 3 systems. Experiments were carried on pressing at room temperature mixtures of powders containing up to 5 wt % of B 4 C or Gd 2 O 3 in Al 2 O 3 and subsequently sintering at 1750 deg C in reducing atmosphere. Evaluation of density, porosity and microstructures were done and a comparison with previous experiences is shown. (Author) [es

  2. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-11-21

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

  3. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-01-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd 2 O 3 ) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

  4. Study and optimization of the composite nuclear fuel with burnable poison UO2/Gd2O3

    International Nuclear Information System (INIS)

    Balestrieri, D.

    1995-09-01

    The studied composite ceramics is a nuclear fuel constituted of a uranium dioxide matrix UO 2 in which big grains (or 'macro-masses') of gadolinium oxide (Gd 2 O 3 ) of 300 ± 100 μm of diameter (mass fraction of 12%) are dispersed. Used as burnable poison (neutron absorbent whose action disappears progressively during the irradiation), gadolinium oxide is the object of a particular attention because some of its properties as the crystal structure, the aptitude to sintering and the thermomechanical behavior have been studied. The aim of this work is to perfect and optimize the process of manufacture of the composite in order to answer to accurate specifications for the density, the shape and the mass fraction of macro-masses. In this framework, it has been necessary to strengthen the Gd 2 O 3 macro-masses by a thermal treatment in order to avoid their deformation during the uniaxial pressing. The influence of this pre-consolidation on the ended microstructure, the aptitude to sintering and the thermal conductivity of the composite have been studied. (O.M.)

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

    International Nuclear Information System (INIS)

    Boyle, R.F.

    1987-01-01

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

  6. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd2O3) with a large number of low concentration gad rods (2 w/o Gd2O3). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (more than 18 months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. This paper also reviews the APA nuclear design code system performance for the low concentration gad design. (author)

  7. Research on application of burnable poison in pebble bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

    Burnable poison in fuel ball was used in pebble bed high-temperature gas-cooled reactor (HTR) to optimize the shape and the peak factor of power distribution in certain conditions. Two options are available and evaluated, that is the homogeneous burnable poison in graphite matrix and burnable poison particles (BPPs) in fuel balls. Due to the absorption cross section of "1"0B, the depletion speed for homogeneous burnable poison is very fast, and difficult to control, on the other side, the depletion speed of BPPs can be optimized respecting to its size, and better shape and peak value of power distribution can be achieved. (authors)

  8. Cutting system for burnable poison rod

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Toyama, Norihide; Koshino, Yasuo; Fujii, Toshio

    1989-01-01

    Burnable poison rods attached to spent fuels are contained in a containing box and transported to a receiving pool. The burnable poison rod-containing box is provisionally situated by the operation to a handling device to a provisional setting rack in a cutting pool and attached to a cutting guide of a cutting device upon cutting. The burnable poison rod is cut only in a cutting pool water and tritium generated upon cutting is dissolved into the cutting pool water. Diffusion of tritium is thus restricted. Further, the cutting pool is isolated by a partition device from the receiving pool during cutting of the burnable poison rod. Accordingly, water in which tritium is dissolved is inhibited from moving to the receiving pool and prevail of tritium contamination can be avoided. (T.M.)

  9. Spider and burnable poison rod combinations

    International Nuclear Information System (INIS)

    Edwards, G.T.; Schluderberg, D.C.

    1980-01-01

    An improved design of burnable poison rods and associated spiders used in fuel assemblies of pressurized water power reactor cores, is described. The rods are joined to the spider arms in a manner which is proof against the reactor core environment and yet allows the removal of the rods from the spider simply, swiftly and delicately. (U.K.)

  10. Spider and burnable poison rod combinations

    International Nuclear Information System (INIS)

    Walton, L.A.

    1980-01-01

    A description is given of an improved design of burnable poison rods and their associated spiders used in the fuel assemblies of pressurized water power reactor cores which allows the rods to be installed and removed more quickly, simply and gently than in previously described systems. (U.K.)

  11. Integrated fuel processor development

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Ferrer, R.M.

    2010-01-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these 'spread' the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  13. Burnable absorber for the PIK reactor

    International Nuclear Information System (INIS)

    Gostev, V.V.; Smolskii, S.L.; Tchmshkyan, D.V.; Zakharov, A.S.; Zvezdkin, V.S.; Konoplev, K.A.

    1998-01-01

    In the reactor PIK design a burnable absorber is not used and the cycle duration is limited by the rods weight. Designed cycle time is two weeks and seams to be not enough for the 100 MW power research reactor equipped by many neutron beams and experimental facilities. Relatively frequent reloading reduces the reactor time on full power and in this way increases the maintenance expenses. In the reactor core fuel elements well mastered by practice are used and its modification was not approved. We try to find the possibilities of installation in the core separate burnable elements to avoid poison of the fuel. It is possible to replace a part of the fuel elements by absorbers, since the fuel elements are relatively small (diameter 5.15mm, uranium 235 content 7.14g) and there are more then 3800 elements in the core. Nevertheless, replacing decreases the fuel burnup and its consumption. In the PIK fuel assembles a little part of the volume is occupied by the dumb elements to create a complete package of the assembles shroud, that is necessary in the hydraulic reasons. In the presented report the assessment of such a replacement is done. As a burnable material Gadolinium was selected. The measurements or the beginning of cycle were performed on the critical facility PIK. The burning calculation was confirmed by measurements on the 18MW reactor WWR-M. The results give the opportunity to twice the cycle duration. The proposed modification of the fuel assembles does not lead to alteration in the other reactor systems, but it touch the burned fuel reprocessing technology. (author)

  14. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    International Nuclear Information System (INIS)

    Grossbeck, M. L.; Renier, J-P.A.; Bigelow, Tim

    2003-01-01

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding

  15. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

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

  16. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1985-01-01

    This patent deals with the fabrication of pellets for neutron absorber rods. Such a pellet includes a matrix of a refractory material which may be aluminum or zirconium oxide, and a burnable poison distributed throughout the matrix. The neutron absorber material may consist of one or more elements or compounds of the metals boron, gadolinium, samarium, cadmium, europium, hafnium, dysprosium and indium. The method of fabricating pellets of these materials outlined in this patent is designed to produce pores or voids in the pellets that can be used to take up the expansion of the burnable poison and to absorb the helium gas generated. In the practice of this invention a slurry of Al 2 O 3 is produced. A hard binder is added and the slurry and binder are spray dried. This powder is mixed with dry B 4 C powder, forming a homogeneous mixture. This mixture is pressed into green tubes which are then sintered. During sintering the binder volatilizes leaving a ceramic with nearly spherical high-density regions of

  17. An evaluation of nuclear design characteristics of duplex burnable poison rods for extended cycle core

    International Nuclear Information System (INIS)

    Lee, D. J.; Kim, M. H.; Song, K. W.

    2003-01-01

    Nuclear design characteristics of duplex burnable poison rod were evaluated for three integral type burnable absorbers; Gadolinia, Erbia and IFBA. Inter-comparison was done for both 12 and 24 month cycle for Korean Standard Nuclear Plant. Fuel assemblies with duplex BP was designed to the equivalent assembly with 8 and 16 gadolinia BP 2 . Duplex BP is composed of inner region of natural U-Gd 2 O 3 , and outer shell of, UO 2 -Er2O 3 . In order to evaluate the duplex BP, assemblies with erbia and IFBA were compared with alternative options. A sensitivity studies were performed to the size of region, compositions and location of duplex BPs. It was shown that duplex BP gave favorable k-infinite curve to burnup, but IFBA provided the least residual reactivity penalty as EOC. Erbia was good for more negative MTCs. IFBA and erbia had better neutronic performance than gadolinia od duplex BP in the aspect of pin power peaking

  18. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1985-01-01

    This invention provides ceramic processing including sintering schedules which produce annular pellets containing burnable poisons for use in reactor control rods. Typically the powder includes Al 2 O 3 and from 1 to 50 weight percent B 4 C. The Al 2 O 3 and B 4 C, appropriately sized, are milled in a ball mill with liquid to produce a slurry. The slurry is spray dried to produce small spheres of the mixed powder, which is mixed with adequate organic binder and plasticizer and formed into a hollow green body having the shape of a tube. The green body is sintered to produce a ceramic tube from which the pellets are cut. The tube is sintered to size so that the pellets have the required dimensions. It is an important feature of this invention that the powder is formed into the green body by applying isostatic pressure to the powder

  19. Burnable neutron absorbers

    International Nuclear Information System (INIS)

    Radford, K.C.; Carlson, W.G.

    1983-01-01

    A neutron-absorber body for use in burnable poison rods in a nuclear reactor. The body is composed of a matrix of Al 2 O 3 containing B 4 C, the neutron absorber. Areas of high density polycrystalline Al 2 O 3 particles are predominantly encircled by pores in some of which there are B 4 C particles. This body is produced by initially spray drying a slurry of A1 2 O 3 powder to which a binder has been added. The powder of agglomerated spheres of the A1 2 O 3 with the binder are dry mixed with B 4 C powder. The mixed powder is formed into a green body by isostatic pressure and the green body is sintered. The sintered body is processed to form the neutron-absorber body. In this case the B 4 C particles are separate from the spheres resulting from the spray drying instead of being embedded in the sphere

  20. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  1. Depletion optimization of lumped burnable poisons in pressurized water reactors

    International Nuclear Information System (INIS)

    Kodah, Z.H.

    1982-01-01

    Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration

  2. Safe core management with burnable absorbers in WWERs

    International Nuclear Information System (INIS)

    1996-01-01

    The objective of this TECDOC is to present state of the art information on burnable poisoned fuel during the CRP. It is based on experimental evidence and on the utilization of theoretical models and will help achieve improvements in safety and economy of LWR cores with hexagonal geometries. 149 refs, figs and tabs

  3. Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corcoran, E.C. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada)], E-mail: emily.corcoran@rmc.ca; Lewis, B.J.; Thompson, W.T. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada); Hood, J. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada); Akbari, F.; He, Z. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ont., K0J 1J0 (Canada); Reid, P. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada)

    2009-03-31

    Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU [CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

  4. Effects of the burnable poison heterogeneity on the long term control of excess of reactivity

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    According to the different geometry shape, the theory of black burnable particles predicts that the evolution of the poison macroscopic absorption cross section is exponentially, quadratic or linear when the burnable poison is displaced in homogeneous distribution, microspheres or needlecylinders heterogeneous distributions, respectively. In the present studies, we took advantage of the Monte Carlo Continuous Energy Burnup Code MCB to verify the black burnable particles theory on the Gas Turbine - Modular Helium Reactor fuelled by military plutonium at the year the fuel reaches the equilibrium composition; we investigated 8 different burnable poisons, B, Cd, Er, Eu, Gd, Dy, Hf and Sm, in three different geometry configurations and we have found that the numerical results qualitatively match the theory predictions when burnable poisons are disposed in small particles

  5. Effects of the burnable poison heterogeneity on the long term control of excess of reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology - KTH, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

    2006-06-15

    According to the different geometry shape, the theory of black burnable particles predicts that the evolution of the poison macroscopic absorption cross section is exponentially, quadratic or linear when the burnable poison is displaced in homogeneous distribution, microspheres or needlecylinders heterogeneous distributions, respectively. In the present studies, we took advantage of the Monte Carlo Continuous Energy Burnup Code MCB to verify the black burnable particles theory on the Gas Turbine - Modular Helium Reactor fuelled by military plutonium at the year the fuel reaches the equilibrium composition; we investigated 8 different burnable poisons, B, Cd, Er, Eu, Gd, Dy, Hf and Sm, in three different geometry configurations and we have found that the numerical results qualitatively match the theory predictions when burnable poisons are disposed in small particles.

  6. Feasibility of using gadolinium as a burnable poison in PWR cores. Final report

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1981-02-01

    As an alternative to the use of lumped burnable absorbers in PWR cores, distributed burnable absorbers are being considered for generic application. These burnable absorbers take the form of Gd 2 O 3 mixed with UO 2 in selected fuel rods (as is currently done in BWR cores). The work discussed herein concerns a three-dimensional feasibility study of the use of such distributed burnable absorbers in PWR cores. This study of distributed burnable absorbers was performed for the first cycle of a typical current design PWR using the following steps: analysis of a generic reference core design; determination of gadolinium assembly designs; determination of a generic gadolinium core design; evaluation of feasibility by examining selected parameters; and redesign of the generic gadolinium core, using axial zoning

  7. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J.A.

    2002-04-17

    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

    1995-01-01

    Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

  9. Neutronic analysis of Gd2O3 as burnable poison

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    For the reactors core design, the use of burnable poisons is one of the options for the control of in excess reactivity and the power form factor. As alternative procedures, the absorbing material may be included in pellets of an inert material or in fuel pellets. Besides, a cladding material and the locations of the fuel elements must be chosen for the first case. The CAREM reactor core design foresees the use of gadolinium oxide (Gd 2 O 3 ) as burnable poison. In this work, a comparative study was made, from the neutronic point of view, among the following alternatives for the poisons location: a) Gd 2 O 3 bars supports in alumina (Al 2 O 3 ), sheathed in steel; b) Gd 2 O 3 bars supports in alumina sheathed in Zry-4; c) Gd 2 O 3 in uranium dioxide (UO 2 ) fuel pellets. (Author) [es

  10. Absorber management using burnable poisons

    International Nuclear Information System (INIS)

    Mortensen, L.

    1977-06-01

    An investigation of the problem of optimal control carried out by means of a two-dimensional model of a PWR reactor. A solution is found to the problem, and the possibility of achieving optimal control with burnable poisons such as boron, cadmium and gadolinium is discussed. Further, an attempt is made to solve the control problem of BWR, but no final solution is found. (author)

  11. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  12. Integral-fuel blocks

    International Nuclear Information System (INIS)

    Cunningham, C.; Simpkin, S.D.

    1975-01-01

    A prismatic moderator block is described which has fuel-containing channels and coolant channels disposed parallel to each other and to edge faces of the block. The coolant channels are arranged in rows on an equilateral triangular lattice pattern and the fuel-containing channels are disposed in a regular lattice pattern with one fuel-containing channel between and equidistant from each of the coolant channels in each group of three mutually adjacent coolant channels. The edge faces of the block are parallel to the rows of coolant channels and the channels nearest to each edge face are disposed in two rows parallel thereto, with one of the rows containing only coolant channels and the other row containing only fuel-containing channels. (Official Gazette)

  13. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR

    International Nuclear Information System (INIS)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G.

    2008-01-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO 2 . (Author)

  14. Neutron evaluation of burnable poison insertion in pressurized water reactor

    International Nuclear Information System (INIS)

    Faria, Rochkhudson Batista de

    2013-01-01

    The development of this work was to match the 'Burn-up Credit Criticality Benchmark - Phase II-D - PWR-UO 2 Assembly Study of Control Rod Effects on Spent Fuel Composition' (case 15), which was modeled using the code MCNP5 and SCALE 6.0. The results of the infinite multiplication factor (k inf ) were compared with those obtained by international institutions. Later we performed in this same benchmark, a sensitivity analysis using SCALE 6.0. Thus, we tested several changes in case 15 of Benchmark, such as insertion of different percentages of burnable poison, changing the number and positions of the rods. In all cases were analyzed, comparisons and discussions about the results. The same methodology was applied to the reactor core of the Nuclear Plant in Brazil, Angra II, initially to evaluate its behavior when subjected to a variation in the percentage of burnable poison and then, introduce changes also in the enrichment of nuclear fuel, doing the appropriate comparisons of results. Considering results and experience gained, the Department of Nuclear Engineering, is prepared to control analysis of reactivity with the use of different types of burnable poisons under the code SCALE 6.0 through its various modules. (author)

  15. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

    2000-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Watanabe, Shoichi; Hirano, Yasushi.

    1998-01-01

    A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

  17. Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

    Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

  18. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Masumi, Ryoji; Ishibashi, Yoko.

    1995-01-01

    A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison-incorporated fuel rods and a spectral shift-type water rod. As the burnable poison for the burnable poison-incorporated fuel rod, a plurality of burnable poison elements each having a different neutron absorption cross section are used. A burnable poison element such as boron having a relatively small neutron absorbing cross section is disposed more in the upper half region than the lower half region of the burnable poison-incorporated fuel rods. In addition, a burnable poison element such as gadolinium having a relatively large neutron absorbing cross section is disposed more in the lower half-region than the upper half region thereof. This can flatten the power distribution in the vertical direction of the fuel assembly and the power distribution in the horizontal direction at the final stage of the operation cycle. (I.N.)

  19. Burnable poison management in light water reactor lattices

    Energy Technology Data Exchange (ETDEWEB)

    Buenemann, D; Mueller, A

    1970-07-01

    For a better reactivity control and power flattening as well as for an increase in dynamic stability the use of burnable poisons in light water reactors has been considered. The main goals for a burnable poison management and its technological realisation are discussed. The poison is assumed to be in the form of separate poison rods or homogeneous or inhomogeneous poisoning in the fuel rods. A new concept with a central poison rod within the fuel rod is discussed. The balance-equation for the needed concentration of burnable poisons for reactivity central as well as the problems of optimization of lumped poisons are treated in connection with the fuel lattice burnup. A first approximation for the design is found. For the calculation of the microburnup of lumped poison and fuel the special code NEUTRA has been developed. The burnup-equation can be chosen either in a simplified burnup-version with 2 pseudo fission products for each fissionable isotope or with an extended system of burnup equations to be used at sophisticated calculations. These burnup equations are coupled to S{sub N}-routines optionally for cylindrical or x-y-geometry for the proper calculation of the microscopic isotope density-, flux-, and power distributions. The theoretical predictions have been checked by means of special experiments so as to determine the accuracy of the computations. Even for a relatively long burnup of the fuel the predicted values are found to be within the experimental error in the case of lumped rods containing a cadmium-alloy or boron carbide. (auth)

  20. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  1. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  2. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi; Masumi, Ryoji; Soneda, Hideo.

    1994-01-01

    A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison rods incorporated with burnable poisons, and water rods which can vary the height in the tube depending on the coolant flow rate flown into the assembly. The amount of entire burnable poisons of the burnable poison-containing rods in adjacent with the water rods is smaller than the amount of entire burnable poisons in the burnable poison containing rods not in adjacent with the water rods. Then the average concentration of burnable poisons in the axial upper half region is made smaller than the average concentration of the burnable poisons at the axial lower half region. Further, a burnable poison concentration at the upper half region of at least one of burnable poison-containing rods in adjacent with the water rods is made lower than the burnable poison concentration in the lower half region. Since this can fasten the combustion of the burnable poisons, a fuel assembly having good fuel economy can be attained. (I.N.)

  3. Low reactivity penalty burnable poison rods

    International Nuclear Information System (INIS)

    1978-01-01

    A nuclear reactor burnable poison rod is described which consists of an elongated tubular sheath enclosing a neutron absorbing material which, at least during reactor operation, also encloses a neutron moderating material. The excess reactivity existing at the beginning of core life is compensated for by the depletion of the burnable poison throughout the life of the core, so that the life of the core is extended. (UK)

  4. Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

    International Nuclear Information System (INIS)

    Edwards, G.T.; Schluderberg, D.C.

    1981-01-01

    A technique is provided for engaging and disengaging burnable poison rods from a spider in a nuclear reactor fuel assembly. A cap on the end of each of the burnable poison rods is provided with a shank or stem that is received in a respective bore formed in the spider. A frangible flange secures the shank and rod to the spider. Pressing the shank in the direction of the bore axis by means, e.g., of a plate ruptures the frangible flange to release the rod from the spider. (author)

  5. Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

    International Nuclear Information System (INIS)

    Walton, L.A.

    1981-01-01

    A method is described of joining burnable poison rods to the spider arms of a pressurised water power reactor fuel assembly which is proof against the reactor core environment but permits these rods to be removed from the spider simply, swiftly and delicately. (U.K.)

  6. Vertical integration in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Mommsen, J.T.

    1977-01-01

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

  7. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

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

  8. Study of burnable poison in the dupic cycle

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da; Almeida, Michel C.B. de; Faria, Rochkhudson B. de; Moreira, Arthur P.C.; Pereira, Claubia, E-mail: clarysson@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

    Recent studies confirm the potential of using reprocessed PWR (Pressurized Water Reactor) fuels in the CANDU (Canada Deuterium Uranium) reactor fuel cycle. An important proposal is the 'Direct Use of spent PWR fuel In CANDU' (DUPIC) cycle, where spent fuels from a PWR are packaged into a CANDU fuel bundle with only mechanical reprocessing (cut into pieces) but no chemical reprocessing. The fissile contents of the spent fuel from Pressurized Water Reactor (PWR) are about 1.5 wt%, which is higher than that of the fuel of CANDU. When this reactor is reload with reprocessed fuel, the reactivity of system will increase and this behavior may affect the safety parameters of reactor. To reduce the initial reactivity, Burnable Poison (BP) can be inserted in the fuel bundle of CANDU. In this way, the present paper evaluates the insertion of the different types of BP considering the DUPIC cycle. The following BPs were evaluated: Boron, Cadmium, Dysprosium, Erbium, Europium, Gadolinium, Hafnium and Samarium. The goal is to verify the neutronic behavior of the fuel bundle at steady state and during the reactor burnup. The SCALE 6.0 (Standardized Computer Analyses for Licensing Evaluation) code was employed to model a standard CANDU-6 fuel element. (author)

  9. Application of B{sub 4}C/Al{sub 2}O{sub 3} Burnable Absorber Rod to Control Excess Reactivity of SMR

    Energy Technology Data Exchange (ETDEWEB)

    Muth, Boravy; Hah, C. J. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    Soluble boron in a nuclear reactor coolant is one of the methods to control excess reactivity of the reactor. However, the use of soluble boron also causes some negative effects such as corrosion, more-positive tendency of Moderator Temperature Coefficient (MTC) and the requirement of Chemical Volume Control System (CVCS). One of the conceptual design features of SMR having been developed in Korea is soluble boron- free reactor to eliminate those drawbacks. Control rods and Burnable Absorber (BA) rods can be other methods than soluble to control excess reactivity. WABA (Wet Annular Burnable Absorber) and PYREX are such type. The other type is IFBA (Integral Fuel Burnable Absorber) in which fuel pellet surface is coated with BA. This paper compares nuclear characteristics of three types of BA as well as SLOBA in terms of k-infinite vs. burnup and explain design basis of SLOBA. This paper also presents the application of SLOBA rods to control long-term excess reactivity of SMR. The SMR loaded with SLOBA rods has been developed for the past few years in Korean. It is named as Bandi-50 with design features of 180 MWth, 37 FAs, fuel assembly height of 200 cm. Soluble-boron-free is one of nuclear design requirements of Bandi-50 and is achieved by controlling excess reactivity of the SMR using BAs and control rods only. To achieve this design requirement, LP is carefully determined in such way that CBC should be as low as possible. Fuel assembly cross-sections are generated by CASMO-3, and core depletion calculations are performed by MASTER.

  10. Irradiation test of borosilicate glass burnable poison

    International Nuclear Information System (INIS)

    Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

    1991-08-01

    The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

  11. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  12. A study on the nuclear characteristics of enriched gadolinia burnable absorber rods; the first year (2000) report

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C.C.; Song, J. S.; Cho, B. O.; Joo, H. G.; Park, S. Y.; Kim, H. Y.; Cho, J. Y.; Kim, K. S.

    2001-04-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

  13. Fuel assemblies

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1983-01-01

    Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

  14. Burnable poison management in a HTR

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, J

    1971-09-21

    It is the purpose with this paper to describe the state-of-the-art of burnable poison investigations made within the Dragon Project and to give the results of a number of calculations, which show that it is possible to control the large initial surplus reactivity of the first core and the radial power distribution with two types of burnable poison sticks with Gadolinium (one type of stick to be used in the inner core region, the other in the outer core region), where the poison will burn away so that keff always stays around the desired value 1.03, and with the radial form-factor not exceeding 1.20. The calculations made for this paper are not too accurate, especially the chosen timestep for calculating the burn-up of the burnable poison stick proved to be too large. Nevertheless, the calculations are good enough to draw the above mentioned conclusions, although they have not given the concentration of Gadolinium to be used in the burnable poison sticks very accurately.

  15. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

    Squires, K.G.

    1995-01-01

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

  16. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

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

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  18. Initial study on burnable poisons in the Dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Pedersen, J

    1971-06-15

    A first study on the effects of burnable poisons in a High Temperature Reactor is given in this paper, and some of the problems concerning the layout and distribution of burnable poison sticks in the core are explained. Time has not allowed us to obtain satisfactory solutions to these problems, but we hope, that this study could form the basis of valuable discussions on ways and means to overcome the difficulties of burnable poison management in HTRs.

  19. Incineration of dry burnable waste from reprocessing plants with the Juelich incineration process

    International Nuclear Information System (INIS)

    Dietrich, H.; Gomoll, H.; Lins, H.

    1987-01-01

    The Juelich incineration process is a two stage controlled air incineration process which has been developed for efficient volume reduction of dry burnable waste of various kinds arising at nuclear facilities. It has also been applied to non nuclear industrial and hospital waste incineration and has recently been selected for the new German Fuel Reprocessing Plant under construction in Wackersdorf, Bavaria, in a modified design

  20. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  1. 46 CFR 119.435 - Integral fuel tanks.

    Science.gov (United States)

    2010-10-01

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

  2. Review of the status of reactor physics predictive methods for burnable poisons in CAGRs

    International Nuclear Information System (INIS)

    Edens, D.J.; McEllin, M.

    1983-01-01

    An essential component of the design of Commercial Advanced Gas Cooled Reactor fuel necessary to achieve higher discharge irradiations is the incorporation of burnable poisons. The poisons enable the more highly enriched fuel required to reach higher irradiation to be loaded without increasing the peak channel power. The optimum choice of fuel enrichment and poison loading will be made using reactor physics predictive methods developed by Berkeley Nuclear Laboratories. These methods and the evidence available to support them from theoretical comparisons, zero energy experiments, WAGR irradiations, and measurements on operating CAGRs are described. (author)

  3. Review of the status of reactor physics predictive methods for burnable poisons in CAGRs

    International Nuclear Information System (INIS)

    Edens, D.J.; McEllin, M.

    1983-01-01

    An essential component of the design of Commercial Advanced Gas Cooled Reactor fuel necessary to achieve higher discharge irradiations is the incorporation of burnable poisons. The poisons enable the more highly enriched fuel required to reach higher irradiation to be loaded without increasing the peak channel power. The optimum choice of fuel enrichment and poison loading will be made using reactor physics predictive methods developed by Berkeley Nuclear Laboratories. The paper describes these methods and the evidence available to support them from theoretical comparisons, zero energy experiments, WAGR irradiations, and measurements on operating CAGR's. (author)

  4. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

    The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  6. Hot channel calculation methodologies in case of Gd burnable poison

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2008-01-01

    The final step in the safety analysis is the investigation of the fulfilment of the acceptance criteria using hot channel calculations. Recently, there has been under way at Paks NPP to introduce a new, higher enriched (4.2 %) fuel type containing Gd burnable poison. To do that, for some transients the DBA analyses must be repeated and last year, as one of the first steps in this process, it was needed to review the hot channel calculation methodologies used in the analyses. The goal of the paper is to summarize some aspects of the hot channel calculation methodologies using different lattice pitches and different fuel types (Gd or non Gd and different enrichments). Mainly, three topics are discussed. First, the influence of the radial power distribution (and other burnup dependent parameters) inside the fuel pin are investigated, and then we discuss the problem of the selection of the appropriate 'frame parameter' in connection with the initial power level at the initial stationary state of DBA transients. Finally, we are trying to answer the question: is it possible to build up a conservative single closed sub-channel approach against multi channel approach?(Authors)

  7. A consolidation process for spent burnable poison rod assemblies

    International Nuclear Information System (INIS)

    Yamamoto, Y.; Harada, M.; Komatsu, Y.

    1985-01-01

    A new consolidation system for the spent burnable poison assembly utilizing a sequence control robot operated under water was proposed. A credible accident in the system was analyzed mainly from the viewpoint of tritium release, based on the diffusion analysis of tritium in borosilicate glass. It was found that the amount of tritium released would be small even after the rupture of burnable poison rods. An experiment on a new consolidation system was performed using spent burnable poison assemblies. The volume of burnable poison assemblies was reduced safely and securely by a factor of 7 to 14 for burnable poison rods and by 22 for hold-down portions. It was proved that the consolidation system is collectively feasible

  8. Rare earths as burnable poison for extended cycles control in electricity generation reactors

    International Nuclear Information System (INIS)

    Asou, M.

    1995-01-01

    The search of an optimization of the French electronuclear network operations leads to a necessary optimization of the core performances. All the economic studies performed by the utilities had shown that there is a real gain to minimize shut down periods for refueling. So, increasing the cycle length from 12 to 18 months will present a gain of shut down for a three years operation period. The theoretical burnable absorber will be a fuel admixed material bringing the required initial negative reactivity with a burn-up kinetic well suited to the fuel and allowing the lowest residual penalty as possible. The residual penalty us defined in this case by the non complete burn up of the poison, by the low of fissile material and by the accumulate of residual isotopes or nuclides. Because of the well known use of gadolinium as burnable absorber for BWR's and PWR's operations, the search for the best compromise to optimize all the above stress is pointed towards the rare earths. In the nuclides family, considering criteria such as cross sections, natural abundance and availability only five nuclides can play the role as burnable absorbers, namely: gadolinium, samarium, dysprosium, europium and erbium. The study presented here will show that only gadolinium and erbium will be considered to control the reactivity of the PWR's. (author). 58 refs., 65 figs., 47 tabs

  9. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  10. Study of the Effect of Burnable Poison Particles Applying in a Pebble Bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing

    2014-01-01

    In pebble bed high temperature gas cooled reactors (HTR), spherical fuel elements pass through the core several times to balance the burnup process in the fuel region, resulting in an acceptable shape and peak factor of power density in the simulation analysis. In contrast, when fuel elements pass through the core only once, the peak of power density occurs at the top of the core and its value is too high to be safe. These indicators/parameters can be improved by incorporating burnable poison in the fuel elements under certain conditions. In the current study, burnable poison particles (BPPs) in fuel elements are evaluated. In spite of the strong absorption capability of "1"0B, BPPs can decrease the depletion speed and increase the duration of "1"0B because of the self-shielding effect, resulting in improved shape and peak factor of power distribution. Several BPPs with different radius are discussed in power distribution, following the calculation for a full-scale reactor core with modified VSOP code. According the result, applying BPPs on fuel pebbles is an effective means to improve the distribution of the power density under one-through fuel load in HTR. (author)

  11. 46 CFR 182.435 - Integral fuel tanks.

    Science.gov (United States)

    2010-10-01

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

  12. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH)3), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an 7 industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  13. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH) 3 ), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  14. Integrating the fuel cycle at IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1992-01-01

    During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

  15. Coal Integrated Gasification Fuel Cell System Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-01-31

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

  16. Optimization of gadolinium burnable poison loading by the conjugate gradients method

    International Nuclear Information System (INIS)

    Drumm, C.R.

    1984-01-01

    Improved use of burnable poison is suggested for pressurized water reactors (PWR's) to insure a sufficiently negative moderator temperature coefficient of reactivity for extended burnup cycles and low leakage refueling patterns. The use of gadolinium as a burnable poison can lead to large axial fluctuations in the power distribution through the cycle. The goal of this work is to determine the optimal axial distribution of gadolinium burnable poison in a PWR to overcome the axial fluctuations, yielding an improved power distribution. The conjugate gradients optimization method is used in this work because of the high degree of nonlinearity of the problem. The neutron diffusion and depletion equations are solved for a one-dimensional one-group core model. The state variables are the flux, the critical soluble boron concentration, and the burnup. The control variables are the number of gadolinium pins per assembly and the beginning-of-cycle gadolinium concentration, which determine the gadolinium cross section. Two separate objectives are considered: 1) to minimize the power peaking factor, which will minimize the capital cost of the plant; and 2) to maximize the cycle length, which will minimize the fuel cost for the plant. It is shown in this work that optimizing the gadolinium distribution can yield an improved power distribution

  17. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Levinskas, D.

    1993-01-01

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

  18. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  19. Assessment of erbium as candidate burnable absorber for future PWR operaning cycles: A neutronic and fabrication study

    International Nuclear Information System (INIS)

    Asou, M.; Dehaudt, P.; Porta, J.

    1995-01-01

    Erbium begins to play a role in the control of PWR core reactivity. Generally speaking, burnable absorbers were only used to establish fresh core equilibrium. In France, since the possibility of extending irradiation cycles by 12 to 18 months, then up to 24 and 30 months, has been envisaged, there is renewed interest in burnable absorbers. The fabrication of PWR pellets has been investigated, providing high density and a good erbium homogeneity. The pellets characteristics were consistent with the specifications of PWR fuel. However, with the present process, the grain size remains small. Studies in progress now shows that erbium is not only a valuable alternative to gadolinium, for long fuel cycles (≥18 months) but also a new fuel concept. (orig.)

  20. Assesment of strength and integrity of fuel channels

    International Nuclear Information System (INIS)

    2000-01-01

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

  1. Integration of fuel cells into residential buildings

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  2. New burnable absorber for long-cycle low boron operation of PWRs

    International Nuclear Information System (INIS)

    Choe, Jiwon; Shin, Ho Cheol; Lee, Deokjung

    2016-01-01

    Highlights: • A burnable absorber design for advanced PWRs with a low soluble boron concentration. • The burnable absorber consists of a UO 2 – 157 Gd 2 O 3 rod with a thin layer of Zr 167 Er 2 . • Three verification cases: two kinds of fuel assemblies and an OPR-1000 core. - Abstract: This paper presents a new high performance burnable absorber (BA) design for advanced Pressurized Water Reactors (PWRs) aiming for a long-cycle operation with a low soluble boron concentration. The new BA consists of a UO 2 – 157 Gd 2 O 3 rod covered with a thin layer of Zr 167 Er 2 . A key feature of this new BA is that enriched isotopes, 157 Gd and 167 Er, are used as absorber materials. Since the high absorption cross section of 157 Gd can reduce the mass fraction of Gd 2 O 3 in UO 2 –Gd 2 O 3 , the thermal margin of fuel rods will increase with higher heat conductivity. Also, the 157 Gd transmutes into 158 Gd by neutron absorption and therefore the residual penalty at the end of cycle (EOC) will decrease. Since 167 Er has a resonance near the thermal neutron energy region, the moderator temperature coefficient (MTC) will become more negative and the control rod worth will increase. These advantages of the new BA are demonstrated with three verification cases: a 17 × 17 Westinghouse (WH) type fuel assembly, a 16 × 16 Combustion Engineering (CE) type fuel assembly, and an OPR-1000 equilibrium core.

  3. Fuel Cell Development and Test Laboratory | Energy Systems Integration

    Science.gov (United States)

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

  4. VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

    International Nuclear Information System (INIS)

    Warner, D.C.; Orr, W.L.

    1985-01-01

    VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

  5. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

    In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

  6. Calculation of burnable cells-Hammer versus Leopard

    International Nuclear Information System (INIS)

    Dias, A.M.; Almeida, C.U.C. de; Pina, C.M. de; Prestes, L.F.; Lederman, L.; Nunes, N.P.; Branco, W.H.

    1977-02-01

    The nuclear parameters for the Angra-1 reactor core are obtained from the cross sections of soluble boron and burnable boron, calculated by the code CITHAM. The results are compared with those developed by the code LEOCIT [pt

  7. Coal Integrated Gasification Fuel Cell System Study

    Energy Technology Data Exchange (ETDEWEB)

    Gregory Wotzak; Chellappa Balan; Faress Rahman; Nguyen Minh

    2003-08-01

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

  8. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

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

  9. Fuel assembly for nuclear reactor

    International Nuclear Information System (INIS)

    Yamanaka, Akihiro; Haikawa, Katsumasa; Haraguchi, Yuko; Nakamura, Mitsuya; Aoyama, Motoo; Koyama, Jun-ichi.

    1996-01-01

    In a BWR type fuel assembly comprising first fuel rods filled with nuclear fission products and second fuel rods filled with burnable poisons and nuclear fission products, the concentration of the burnable poisons mixed to a portion of the second fuel rods is controlled so that it is reduced at the upper portion and increased at the lower portion in the axial direction. In addition, a product of the difference of an average concentration of burnable poisons between the upper portion and the lower portion and the number of fuel rods is determined to higher than a first set value determined corresponding to the limit value of a maximum linear power density. The sum of the difference of the average concentration of the burnable poisons between the upper portion and the lower portion of the second fuel rod and the number of the second fuel rods is determined to lower than a second set value determined corresponding to a required value of a surplus reactivity. If the number of the fuel rods mixed with the burnable poisons is increased, the infinite multiplication factor at an initial stage of the burning is lowered and, if the concentration of the mixed burnable poisons is increased, the time of exhaustion of the burnable poisons is delayed. As a result, the maximum value of the infinite multiplication factor is suppressed thereby enabling to control surplus reactivity. (N.H.)

  10. Integrated international safeguards concepts for fuel reprocessing

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Gutmacher, R.G.; Markin, J.T.; Shipley, J.P.; Whitty, W.J.; Camp, A.L.; Cameron, C.P.; Bleck, M.E.; Ellwein, L.B.

    1981-12-01

    This report is the fourth in a series of efforts by the Los Alamos National Laboratory and Sandia National Laboratories, Albuquerque, to identify problems and propose solutions for international safeguarding of light-water reactor spent-fuel reprocessing plants. Problem areas for international safeguards were identified in a previous Problem Statement (LA-7551-MS/SAND79-0108). Accounting concepts that could be verified internationally were presented in a subsequent study (LA-8042). Concepts for containment/surveillance were presented, conceptual designs were developed, and the effectiveness of these designs was evaluated in a companion study (SAND80-0160). The report discusses the coordination of nuclear materials accounting and containment/surveillance concepts in an effort to define an effective integrated safeguards system. The Allied-General Nuclear Services fuels reprocessing plant at Barnwell, South Carolina, was used as the reference facility

  11. Fuel tank integrity research : fuel tank analyses and test plans

    Science.gov (United States)

    2013-04-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

  12. the effect of advanced fuel designs on fuel utilization

    International Nuclear Information System (INIS)

    Sarikaya, B.; Colak, U.; Tombakoglu, M.; Yilmazbayhan, A.

    1997-01-01

    Fuel management is one of the key topic in nuclear engineering. It is possible to increase fuel burnup and reactor lifetime by using advanced fuel management strategies. In order to increase the cycle lifetime, required amount of excess reactivity must be added to system. Burnable poisons can be used to compensate this excess reactivity. Usually gadolinium (Gd) is used as burnable poison. But the use of Gd presents some difficulties that have not been encountered with the use of boron

  13. Calculation of resonance integral for fuel cluster

    International Nuclear Information System (INIS)

    Remsak, S.

    1969-01-01

    The procedure for calculating the shielding correction, formulated in the previous paper [6], was broadened and applied for a cluster of cylindrical rods. The sam analytical method as in the previous paper was applied. A combination of Gauss method with the method of Almgren and Porn used for solving the same type of integral was used to calculate the geometry functions. CLUSTER code was written for ZUSE-Z-23 computer to calculate the shielding corrections for pairs of fuel rods in the cluster. Computing time for one pair of fuel rods depends on the number of closely placed rod, and for two closely placed rods it is about 3 hours. Calculations were done for clusters containing 7 and 19 UO 2 rods. results show that calculated values of resonance integrals are somewhat higher than the values obtained by Helstrand empirical formula. Taking into account the results for two rods from the previous paper it can be noted that the calculated and empirical values for clusters with 2 and 7 rods are in agreement since the deviations do not exceed the limits of experimental error (±2%). In case of larger cluster with 19 rods deviations are higher than the experimental error. Most probably the calculated values exceed the experimental ones result from the fact that in this paper the shielding correction is calculated only in the region up to 1 keV [sr

  14. Preliminary study or RSG-GAS reactor fuel element integrity

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Evaluation of the in pile performance of boron containing fuel pellets

    International Nuclear Information System (INIS)

    Jeong, Gwanyoon; Sohn, Dongseong

    2012-01-01

    The world rare earth resource are heavily concentrated in certain area and if these natural resources are weaponized by a country, we may confront serious difficulty because rare earth element gadolinium(Gd) is used as burnable poison material in some nuclear power plants (NPP) in Korea. Gd is used as a neutron absorbing material in Gd 2 O 3 form and mixed with UO 2 When boron is used as burnable poison in nuclear fuel, in fuel pellets. The burnable poison mixed in the fuel pellets is called integral burnable absorber (BA) design which differentiates it from the old separate BA design. In the old separate BA design, boron(B) was used in borosilicate glass (PYREX) form and placed in guide tubes. With the development of the concern over the availability of rare earth material Gd, B is considered as a candidate material replacing Gd for the case when the rare earth material is weaponized. However the idea for new boron BA design is integral type because the integral type BA design has several benefits over the separate BA design, such as reduction of radioactive waste, more positions for BA location, etc. 10 B absorbs a neutron and produces helium by the following reaction: 10 B + n → 7 Li + 4 He The helium produced by the nuclear reaction may cause the increase of rod internal pressure and change the gap conductivity if the significant amount of helium gas is released to the gap between the pellet and the cladding. Thus, it is necessary to investigate the in-pile behaviors of B containing pellet. However, few experiment have been carried out so far on the behavior of in-pile produced helium in UO 2 fuel pellets, especially for the cases boron compound is mixed with UO 2 In this paper, we will evaluate the production and the release of helium depending on fuel. 10 B concentration in the fuel

  16. Integrated data base for spent fuel and radwaste: inventories

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  17. Experimental and theoretical burnup investigations on model arrangements with solid burnable poisons

    International Nuclear Information System (INIS)

    Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

    1975-01-01

    It is the scope of the two experiments here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

  18. Experimental and theoretical investigations on solid burnable poison burnup of model arrangements

    International Nuclear Information System (INIS)

    Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

    1975-01-01

    It is the scope of the two experiments reported here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

  19. Heterogeneous burnable poisons. Sinterability study in oxidizing atmosphere of alumina-gadolinia and alumina-boron carbide compounds

    International Nuclear Information System (INIS)

    Agueda, H.C.; Leiva, S.F.; Russo, D.O.

    1990-01-01

    Solid burnable poisons are used in reactors cooled by pressure light water (PLWR) with the purpose of controlling initial reactivity in the first reactor's core. The burnable poisons may be uniformly mixed with the fuel -known as 'homogeneous' poisons-; or constituting separate elements -known as heterogeneous poisons-. The purpose of this work is to present the results of two sinterability studies, performed on Al 2 O 3 -Gd 2 O 3 and Al 2 O 3 -B 4 C, where alumina acts as inert matrix, storing the absorbing elements as Gd 2 O 3 or B 4 C. The elements were sintered at an air atmosphere and additives permitting the obtention of a greater density alumina were tested at lower temperatures than the characteristic for this material, in order to determine its compatibility with the materials dealt with herein. (Author) [es

  20. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

    Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

  1. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  2. Integrity of spent CANDU fuel during and following dry storage

    International Nuclear Information System (INIS)

    Villagran, J.E.

    2004-01-01

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

  3. Burnable poison calculations for Mk.III gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Gubbins, M E

    1971-02-15

    A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.

  4. Study of low leakage reload schedulle without burnable posion for Angra-1

    International Nuclear Information System (INIS)

    Sakai, M.; Dias, A.

    1989-01-01

    At the moment, there is a world trend to design larger cycles for PWR. Then the reload batches are increased, the enrichment in 235 U is increased and/or advanced fuel management strategies with radial low neutron leakage are applied. For the low leakage reloads of Angra-1 calculations were performed for different number of fuel assemblies for reaload batch, 32,36,40,44 and 48, from the 4th cycle up to equilibrium cycle for two different enrichments 3,4 W/O and 3,9 W/O in 235 U. The results showed that for the enrichments used without burnable posion it is possible to reach an increase in cycle lenghts between 3% and 8% for the same conditions. (author) [pt

  5. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167 Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1 B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167 Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff , at the beginning of the operation

  6. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

    2006-01-15

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the {sup 167}Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in {sup 1}B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k {sub eff} value, the {sup 167}Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the {beta} {sub eff}, at the beginning of the operation.

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

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

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

  8. Nuclear Fuel Design Considerations for the 1990s

    International Nuclear Information System (INIS)

    Stucker, David L.

    1993-01-01

    materials, minimum residual burnable absorbers and optimized fuel assembly designs customized to the specific needs of an individual utility and/or power reactor. The advances of the fabricators have been made possible, in large part, by improvements in vendor/utility communication as well as closely-integrated technology and hardware solutions to the competitive demands of the industry. The dynamics of the nuclear fuel supply relationship between electric utilities and nuclear fuel suppliers has evolved over the years from a clear cut supplier/customer relationship in the early years to today's relationships characterized by technical partnerships.

  9. An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Erich [Univ. of Texas, Austin, TX (United States); Scopatz, Anthony [Univ. of Wisconsin, Madison, WI (United States)

    2016-04-25

    Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

  10. Low NOx Fuel Flexible Combustor Integration Project Overview

    Science.gov (United States)

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

    2015-01-01

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

  11. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  12. Safeguards operations in the integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  13. Preparations for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    International Nuclear Information System (INIS)

    Papukchiev, Angel; Schaefer, Anselm

    2008-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  15. ABB Turbo advanced fuel for application in System 80 family of plants

    International Nuclear Information System (INIS)

    Karoutas, Z.E.; Dixon, D.J.; Shapiro, N.L.

    1998-01-01

    ABB Combustion Engineering Nuclear Operations (ABB CE) has developed an Advanced Fuel Design, tailored to the Combustion Engineering, Inc. (CE) Nuclear Steam Supply System (NSSS) environment. This Advanced Fuel Design called Turbo features a full complement of innovative components, including GUARDIAN debris-resistant spacer grids, Turbo Zircaloy mixing grids to increase thermal margin and grid-to-rod fretting resistance, value-added fuel pellets to increase fuel loading, advanced cladding to increase achievable burnup, and axial blankets and Erbium integral burnable absorbers for improving fuel cycle economics. This paper summarizes the Turbo Fuel Design and its application to a System 80 family type plant. Benefits in fuel reliability, thermal margin, improved fuel cycle economics and burn up capability are compared relative to the current ABB CE standard fuel design. The fuel management design and the associated thermal margin are also evaluated. (author)

  16. The integrity of CANDU fuel during load following

    International Nuclear Information System (INIS)

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

    1989-08-01

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

  17. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

    1996-01-01

    In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

  18. Nuclear fuel clad clothed with burnable poison and obtainment process

    International Nuclear Information System (INIS)

    Diez, P.; Netter, P.

    1994-01-01

    This clad has preferentially on its inner surface a boron compound such boron carbide or boron nitrogen deposited by Chemical Vapor Deposition or by Physical Vapor Deposition without any temperature elevation injurious to its mechanical properties. 3 figs

  19. Feasibility study of the design of homogeneously mixed thorium-uranium oxide and all-uranium fueled reactor cores for civil nuclear marine propulsion - 15082

    International Nuclear Information System (INIS)

    Alam, S.B.; Lindley, B.A.; Parks, G.T.

    2015-01-01

    In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)

  20. Genetic algorithm for the optimization of the loading pattern for reactor core fuel management

    International Nuclear Information System (INIS)

    Zhou Sheng; Hu Yongming; zheng Wenxiang

    2000-01-01

    The paper discusses the application of a genetic algorithm to the optimization of the loading pattern for in-core fuel management with the NP characteristics. The algorithm develops a matrix model for the fuel assembly loading pattern. The burnable poisons matrix was assigned randomly considering the distributed nature of the poisons. A method based on the traveling salesman problem was used to solve the problem. A integrated code for in-core fuel management was formed by combining this code with a reactor physics code

  1. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  2. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

  3. Integrated fuel-cycle models for fast breeder reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. Calculation qualification of gadolinium burnable poisons in water reactors

    International Nuclear Information System (INIS)

    Chaucheprat, P.

    1988-01-01

    The work presented in this thesis constitutes the qualification on the one end of Appolo-Neptune scheme for the gadolinium burnable poison in a pressurized water reactor, and on the other end of basis nuclear data on natural gadolinium. This study has permitted to reduce by a factor 3 the actual incertitude on the gadolinium poison comparatively at precisions cited in international benchmarks calculations [fr

  5. Gadolinium burnable absorber optimization by the method of conjugate gradients

    International Nuclear Information System (INIS)

    Drumm, C.R.; Lee, J.C.

    1987-01-01

    The optimal axial distribution of gadolinium burnable poison in a pressurized water reactor is determined to yield an improved power distribution. The optimization scheme is based on Pontryagin's maximum principle, with the objective function accounting for a target power distribution. The conjugate gradients optimization method is used to solve the resulting Euler-Lagrange equations iteratively, efficiently handling the high degree of nonlinearity of the problem

  6. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  7. Fuel conservation integrated into airline economics

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, D.R.

    1981-01-01

    Fuel conservation efforts at most major airlines involve close scrutiny and intensive analysis in all areas - flight, maintenance and ground handling. Yet, despite the concern and attention devoted, the fundamental question of fuel saving versus time trade-offs remains unanswered. This paper introduces and defines the concept ''The value of an airplane to an airline is that airplane's earning power.

  8. Apparatus for integrated fuel assembly inspection system

    International Nuclear Information System (INIS)

    Ahmed, H.J.; Burchill, S.R.

    1988-01-01

    In a fuel assembly inspection apparatus, the combination is described comprising: (a) an elongated fixture mounted in a stationary upright position; (b) upper means mounted to an upper portion of the fixture and lower means mounted adjacent to a lower portion of the fixture, the upper and lower means being disposed outwardly from a side of the fixture for supporting a nuclear fuel assembly therebetween and extending along the side of the fixture; (c) a bottom carriage having a central opening adapted to receive the fuel assembly therethrough when supported between the upper and lower means such that the bottom carriage being connected only to, and extending in cantilever fashion outwardly from, the side of the fixture for generally vertical movement along the side of the fixture and along the fuel assembly extending along the side of the fixture; (d) drive means for selectively moving the bottom carriage; and (e) means disposed on the bottom carriage for measuring the envelop, of the fuel assembly when the bottom carriage is moved to and stationed at selected axial positions along the fuel assembly

  9. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    International Nuclear Information System (INIS)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor

  10. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Walters, L.C.

    1985-01-01

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

  11. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  12. The research on burnup characteristic of doping burnable poison in PWR

    International Nuclear Information System (INIS)

    Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

    2014-01-01

    In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

  13. Burnable absorber rod releasable latching structure

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Wilson, J.F.

    1987-01-01

    An elongated nuclear reactivity control member, a releasable latching structure useful for releasably attaching the control member at an end to a top nozzle adapter plate of a nuclear fuel assembly is described comprising: (a) a mounting body including an inner plug portion attached to the end of the control member and an outer end portion disposed axially outward from the inner plug portion and the end of the member; and (b) a spring latch disposed about the mounting body and being attached to the outer end portion. The spring latch has at least one latch finger extending toward the inner plug portion of the body and is movable toward and away from the body between an outer latching position in which the finger is adapted to engage a fuel assembly top nozzle adapter plate and retain the elongated member in a stationary relationship with respect to the adapter plate and an inner unlatching position in which the finger is adapted to disengage from the adapter plate and allow removal of the member from the adapter plate

  14. Industrial integration of the fuel cycle in Argentina

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  15. Industrial integration of the fuel cycle in Argentina

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  17. Results of trial operation of the WWER advanced fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

    2001-01-01

    The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

  18. Integrated analysis of oxide nuclear fuel sintering

    International Nuclear Information System (INIS)

    Baranov, V.; Kuzmin, R.; Tenishev, A.; Timoshin, I.; Khlunov, A.; Ivanov, A.; Petrov, I.

    2011-01-01

    Dilatometric and thermal-gravimetric investigations have been carried out for the sintering process of oxide nuclear fuel in gaseous Ar - 8% H 2 atmosphere at temperatures up to 1600 0 C. The pressed compacts were fabricated under real production conditions of the OAO MSZ with application of two different technologies, so called 'dry' and 'wet' technologies. Effects of the grain size growth after the heating to different temperatures were observed. In order to investigate the effects produced by rate of heating on properties of sintered fuel pellets, the heating rates were varied from 1 to 8 0 C per minute. Time of isothermal overexposure at maximal temperature (1600 0 C) was about 8 hours. Real production conditions were imitated. The results showed that the sintering process of the fuel pellets produced by two technologies differs. The samples sintered under different heating rates were studied with application of scanning electronic microscopy analysis for determination of mean grain size. A simulation of heating profile for industrial furnaces was performed to reduce the beam cycles and estimate the effects of variation of the isothermal overexposure temperatures. Based on this data, an optimization of the sintering conditions was performed in operations terms of OAO MSZ. (authors)

  19. Waste management in IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Integral benchmarks with reference to thorium fuel cycle

    International Nuclear Information System (INIS)

    Ganesan, S.

    2003-01-01

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

  1. Description of reactor fuel breeding with three integral concepts

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  2. Implementation of strength pareto evolutionary algorithm II in the multiobjective burnable poison placement optimization of KWU pressurized water reactor

    International Nuclear Information System (INIS)

    Gharari, Rahman; Poursalehi, Navid; Abbasi, Mohmmadreza; Aghale, Mahdi

    2016-01-01

    In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (K-e-f-f) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor

  3. Implementation of strength pareto evolutionary algorithm II in the multiobjective burnable poison placement optimization of KWU pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gharari, Rahman [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Poursalehi, Navid; Abbasi, Mohmmadreza; Aghale, Mahdi [Nuclear Engineering Dept, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (K-e-f-f) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

  4. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  5. Integrated fuel cell stack shunt current prevention arrangement

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  6. Oxy-fuel combustion with integrated pollution control

    Science.gov (United States)

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

    2012-01-03

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

  7. The FIT Model - Fuel-cycle Integration and Tradeoffs

    International Nuclear Information System (INIS)

    Piet, Steven J.; Soelberg, Nick R.; Bays, Samuel E.; Pereira, Candido; Pincock, Layne F.; Shaber, Eric L.; Teague, Melissa C.; Teske, Gregory M.; Vedros, Kurt G.

    2010-01-01

    All mass streams from fuel separation and fabrication are products that must meet some set of product criteria - fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the 'system losses study' team that developed it (Shropshire2009, Piet2010) are an initial step by the FCR and D program toward a global analysis that accounts for the requirements and capabilities of each component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R and D needs and set longer-term goals. The question originally posed to the 'system losses study' was the cost of separation, fuel fabrication, waste management, etc. versus the separation efficiency. In other words, are the costs associated with marginal reductions in separations losses (or improvements in product recovery) justified by the gains in the performance of other systems? We have learned that that is the wrong question. The right question is: how does one adjust the compositions and quantities of all mass streams, given uncertain product criteria, to balance competing objectives including cost? FIT is a method to analyze different fuel cycles using common bases to determine how chemical performance changes in one part of a fuel cycle (say used fuel cooling times or separation efficiencies) affect other parts of the fuel cycle. FIT estimates impurities in fuel and waste via a rough estimate of physics and mass balance for a set of technologies. If feasibility is an issue for a set, as it is for 'minimum fuel treatment' approaches such as melt refining and AIROX, it can help to make an estimate of how performances would have to change to achieve feasibility.

  8. PHWR Fuel - an integrated approach in Indian context

    Energy Technology Data Exchange (ETDEWEB)

    Jayaraj, R.N. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderabad (India)

    2008-07-01

    The nuclear power programme in India is based on a three-stage approach in which the Pressurized Heavy Water Reactors (PHWR) forms the backbone of the first stage. Over the years, apart from gaining expertise in design, construction and operation of PHWRs, innovative fuel designs and manufacturing technologies have also been evolved. Presently, thirteen PHWR 220 units and two PHWR 540 units are in operation. Three more PHWR 220 units are in the advanced stage of construction. In addition, the PHWR power generation programme envisages construction of eight more PHWR 700 units. Nuclear Fuel Complex (NFC) at Hyderabad, established in early 70s, is the only manufacturer of fuel and reactor core structurals for all the PHWRs in India. Since inception, the thrust has been on indigenous development of technology in the areas of production processes, equipment manufacture and quality assurance programmes. Commensurate with the PHWR programme, NFC has expanded its production capacities and has fabricated more than 380,000 fuel bundles since inception. Towards optimization of uranium resources and implementation of 'closed fuel cycle' concept, large quantities of reprocessed uranium fuel bundles have been manufactured and introduced in the initial cores of PHWRs. In recent times, NFC introduced several modifications in the production processes like vapour ammonia precipitation for UO{sub 2} powder production, advanced resistance welding controls and improved versions of welding machines, which all have facilitated in improving the quality and productivity of the fuel. Superior quality control systems like spectrophotometric determination of SSA of UO{sub 2} powders, machine vision systems for pellet inspection, thermography for evaluating weld integrity, etc. has channelised NDT techniques into fuel production lines. The paper summarizes various improvements carried out in the design and manufacture of PHWR fuel. New concepts evolved in high burn-up fuels and

  9. PHWR Fuel - an integrated approach in Indian context

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2008-01-01

    The nuclear power programme in India is based on a three-stage approach in which the Pressurized Heavy Water Reactors (PHWR) forms the backbone of the first stage. Over the years, apart from gaining expertise in design, construction and operation of PHWRs, innovative fuel designs and manufacturing technologies have also been evolved. Presently, thirteen PHWR 220 units and two PHWR 540 units are in operation. Three more PHWR 220 units are in the advanced stage of construction. In addition, the PHWR power generation programme envisages construction of eight more PHWR 700 units. Nuclear Fuel Complex (NFC) at Hyderabad, established in early 70s, is the only manufacturer of fuel and reactor core structurals for all the PHWRs in India. Since inception, the thrust has been on indigenous development of technology in the areas of production processes, equipment manufacture and quality assurance programmes. Commensurate with the PHWR programme, NFC has expanded its production capacities and has fabricated more than 380,000 fuel bundles since inception. Towards optimization of uranium resources and implementation of 'closed fuel cycle' concept, large quantities of reprocessed uranium fuel bundles have been manufactured and introduced in the initial cores of PHWRs. In recent times, NFC introduced several modifications in the production processes like vapour ammonia precipitation for UO 2 powder production, advanced resistance welding controls and improved versions of welding machines, which all have facilitated in improving the quality and productivity of the fuel. Superior quality control systems like spectrophotometric determination of SSA of UO 2 powders, machine vision systems for pellet inspection, thermography for evaluating weld integrity, etc. has channelised NDT techniques into fuel production lines. The paper summarizes various improvements carried out in the design and manufacture of PHWR fuel. New concepts evolved in high burn-up fuels and development of state

  10. Integrating fuel cell power systems into building physical plants

    Energy Technology Data Exchange (ETDEWEB)

    Carson, J. [KCI Technologies, Inc., Hunt Valley, MD (United States)

    1996-12-31

    This paper discusses the integration of fuel cell power plants and absorption chillers to cogenerate chilled water or hot water/steam for all weather air conditioning as one possible approach to building system applications. Absorption chillers utilize thermal energy in an absorption based cycle to chill water. It is feasible to use waste heat from fuel cells to provide hydronic heating and cooling. Performance regimes will vary as a function of the supply and quality of waste heat. Respective performance characteristics of fuel cells, absorption chillers and air conditioning systems will define relationships between thermal and electrical load capacities for the combined systems. Specifically, this paper develops thermodynamic relationships between bulk electrical power and cooling/heating capacities for combined fuel cell and absorption chiller system in building applications.

  11. Spray sealing: A breakthrough in integral fuel tank sealing technology

    Science.gov (United States)

    Richardson, Martin D.; Zadarnowski, J. H.

    1989-11-01

    In a continuing effort to increase readiness, a new approach to sealing integral fuel tanks is being developed. The technique seals potential leak sources by spraying elastomeric materials inside the tank cavity. Laboratory evaluations project an increase in aircraft supportability and reliability, an improved maintainability, decreasing acquisition and life cycle costs. Increased usable fuel volume and lower weight than conventional bladders improve performance. Concept feasibility was demonstrated on sub-scale aircraft fuel tanks. Materials were selected by testing sprayable elastomers in a fuel tank environment. Chemical stability, mechanical properties, and dynamic durability of the elastomer are being evaluated at the laboratory level and in sub-scale and full scale aircraft component fatigue tests. The self sealing capability of sprayable materials is also under development. Ballistic tests show an improved aircraft survivability, due in part to the elastomer's mechanical properties and its ability to damp vibrations. New application equipment, system removal, and repair methods are being investigated.

  12. Economic competitiveness of fuel cell onsite integrated energy systems

    Science.gov (United States)

    Bollenbacher, G.

    1983-01-01

    The economic competitiveness of fuel cell onsite integrated energy systems (OS/IES) in residential and commercial buildings is examined. The analysis is carried out for three different buildings with each building assumed to be at three geographic locations spanning a range of climatic conditions. Numerous design options and operating strategies are evaluated and two economic criteria are used to measure economic performance. In general the results show that fuel cell OS/IES's are competitive in most regions of the country if the OS/IES is properly designed. The preferred design is grid connected, makes effective use of the fuel cell's thermal output, and has a fuel cell powerplant sized for the building's base electrical load.

  13. 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

  14. Fuel pin integrity assessment under large scale transients

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2006-01-01

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

  15. Integrating Wind And Solar With Hydrogen Producing Fuel Cells

    NARCIS (Netherlands)

    Hemmes, K.

    2007-01-01

    The often proposed solution for the fluctuating wind energy supply is the conversion of the surplus of wind energy into hydrogen by means of electrolysis. In this paper a patented alternative is proposed consisting of the integration of wind turbines with internal reforming fuel-cells, capable of

  16. Integral reactor system and method for fuel cells

    Science.gov (United States)

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  17. Determining fissile content of nuclear fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

    1980-01-01

    This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

  18. New Small LWR Core Designs using Particle Burnable Poisons for Low Boron Concentration

    International Nuclear Information System (INIS)

    Yoo, Ho Seong; Hwang, Dae Hee; Hong, Ser Gi

    2015-01-01

    The soluble boron has two major important roles in commercial PWR operations : 1) the control of the long-term reactivity to maintain criticality under normal operation, and 2) the shutdown of the reactor under accidents. However, the removal of the soluble boron gives several advantages in SMRs (Small Modular Reactor). These advantages resulted from the elimination of soluble boron include the significant simplification of nuclear power plant through the removal of pipes, pumps, and purification systems. Also, the use of soluble boron mitigates corrosion problems on the primary coolant loop. Furthermore, the soluble boron-free operation can remove an inadvertent boron dilution accident (BDA) which can lead to a significant insertion of positive reactivity. From the viewpoint of core physics, the removal of soluble boron or reduction of soluble boron concentration makes the moderator temperature coefficient (MTC) more negative. From the core design studies using new fuel assemblies, it is shown that the cores have very low critical soluble boron concentrations less than 500ppm, low peaking factors within the design targets, strong negative MTCs over cycles, and large enough shutdown margins both at BOC and EOC. However, the present cores have relatively low average discharge burnups of ∼ 30MWD/kg leading to low fuel economy because the cores use lots of non-fuel burnable poison rods to achieve very low critical boron concentrations. So, in the future, we will perform the trade-off study between the fuel discharge burnup and the boron concentrations by changing fuel assembly design and the core loading pattern

  19. New Small LWR Core Designs using Particle Burnable Poisons for Low Boron Concentration

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Ho Seong; Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The soluble boron has two major important roles in commercial PWR operations : 1) the control of the long-term reactivity to maintain criticality under normal operation, and 2) the shutdown of the reactor under accidents. However, the removal of the soluble boron gives several advantages in SMRs (Small Modular Reactor). These advantages resulted from the elimination of soluble boron include the significant simplification of nuclear power plant through the removal of pipes, pumps, and purification systems. Also, the use of soluble boron mitigates corrosion problems on the primary coolant loop. Furthermore, the soluble boron-free operation can remove an inadvertent boron dilution accident (BDA) which can lead to a significant insertion of positive reactivity. From the viewpoint of core physics, the removal of soluble boron or reduction of soluble boron concentration makes the moderator temperature coefficient (MTC) more negative. From the core design studies using new fuel assemblies, it is shown that the cores have very low critical soluble boron concentrations less than 500ppm, low peaking factors within the design targets, strong negative MTCs over cycles, and large enough shutdown margins both at BOC and EOC. However, the present cores have relatively low average discharge burnups of ∼ 30MWD/kg leading to low fuel economy because the cores use lots of non-fuel burnable poison rods to achieve very low critical boron concentrations. So, in the future, we will perform the trade-off study between the fuel discharge burnup and the boron concentrations by changing fuel assembly design and the core loading pattern.

  20. Generalized pin factor methodology for LWR reload cores with discrete burnable absorbers

    International Nuclear Information System (INIS)

    Hah, C.J.; Hideki Matsumoto; Toshikazu Ida; Lee, C.; Chao, Y.A.

    2005-01-01

    Discrete burnable absorbers are used to suppress excess reactivity as well as peak pin power in an assembly. After the burn-out of absorption material, discrete burnable absorbers are usually removed from assembly guide tubes for the next cycle. For that case, the pin factors with discrete burnable absorbers cannot be used since the assembly configuration is physically changed. The pin factors without discrete burnable absorbers also have noticeable deviation from the actual case because they do not take into account the history effect due to the residence of discrete burnable absorbers for the previous cycle. In this paper, the generalized pin factor (GPF) method is developed to accurately predict pin powers by considering the history effect. The method uses a second-order polynomial function to approximate the history effect which builds up during the residence of burnable absorber material and employs a linear approximation to simulate the decay of the history effect after discrete burnable absorbers are removed. The verification results from Westinghouse Vantage- 5H assemblies with WABAs showed that pin power errors were significantly reduced by using the GPF. (authors)

  1. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  2. Special topics of inner fuel management

    International Nuclear Information System (INIS)

    Wuenschmann, A.

    1977-01-01

    Burnable Poison Rod Assemblies (BPRA) are currently used as lumped burnable poison only in the first cycles of many power reactors to insure a negative moderator coefficient at beginning of life and to help shape core power distribution (out-in shuffle scheme). BPRA's are also a valuable tool in later cycles where they can be used as an additional design parameter to improve fuel performance and fuel cycle economics, to shape fuel assembly power, and to increase fuel management flexibility (in-out shuffle scheme). This paper describes the two fuel shuffle schemes and compares the two shuffle strategies concerning economic and flexibility aspects. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

  4. Integrated fuel cell energy system for modern buildings

    Energy Technology Data Exchange (ETDEWEB)

    Moard, D.M.; Cuzens, J.E.

    1998-07-01

    Energy deregulation, building design efficiency standards and competitive pressures all encourage the incorporation of distributed fuel cell cogeneration packages into modern buildings. The building marketplace segments to which these systems apply include office buildings, retail stores, hospitals, hotels, food service and multifamily residences. These applications represent approximately 60% of the commercial building sector's energy use plus a portion of the residential sector's energy use. While there are several potential manufacturers of fuel cells on the verge of marketing equipment, most are currently using commercial hydrogen gas to fuel them. There are few suppliers of equipment, which convert conventional fuels into hydrogen. Hydrogen Burner Technology, Inc. (HBT) is one of the few companies with a proven under-oxidized-burner (UOB) technology, patented and already proven in commercial use for industrial applications. HBT is developing a subsystem based on the UOB technology that can produce a hydrogen rich product gas using natural gas, propane or liquid fuels as the feed stock, which may be directly useable by proton exchange membrane (PEM) fuel cells for conversion into electricity. The combined thermal output can also be used for space heating/cooling, water heating or steam generation applications. HBT is currently analyzing the commercial building market, integrated system designs and marketplace motivations which will allow the best overall subsystem to be designed, tested and introduced commercially in the shortest time possible. HBT is also actively involved in combined subsystem designs for use in automotive and small residential services.

  5. Integrated risk assessment for spent fuel transportation using developed software

    International Nuclear Information System (INIS)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun; Lee, Sang hoon

    2016-01-01

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed

  6. Integrated risk assessment for spent fuel transportation using developed software

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun [KAIST, Daejeon (Korea, Republic of); Lee, Sang hoon [Keimyung University, Daegu (Korea, Republic of)

    2016-05-15

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed.

  7. Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

    International Nuclear Information System (INIS)

    Huang Jinhua; Sun Hanhong

    1988-08-01

    Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  9. First results on study of gadolinium as burnable absorber

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    Following on with the work included in the 'Burnable absorbers research plan' several experiments were carried out oriented to determine Ga 2 O 3 burn up. Cold tests were performed and samples were irradiated in the RA-3 reactor. In this paper, some calculated values are presented together with their comparisons with experimental ones. The parameters foreseen for performing the experiments were verified and also the predictions on burn up of uranium and gadolinium isotopes concentrations. These results imply that the nuclear data of these isotopes included in the library are satisfactory. Next steps will be to measure other isotopes concentrations, gamma spectrum, and the irradiation of one pellet to determine self shielding effects in order to obtain effective cross sections i.e. for CAREM geometry. (author)

  10. The treatment of burnable poison pins in LWRWIMS

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1982-12-01

    This report describes an investigation into the modelling approximations normally made when the LWR lattice code LWRWIMS is used for design calculations on assemblies containing burnable poison pins. Parameters investigated include energy group structure, intervals between calculations in MWd/te and spatial subdivision of the poison pins. An estimate is made of the effect of using pin-cell smearing with diffusion theory for the assembly geometry, instead of a more exact heterogeneous transport theory calculation. The influence on reactivity of the minor gadolinium isotopes 152, 154, 156, 158 and 160 in a poison pin dominated by the isotopes 155 and 157 is presented, and finally, recommendations on the use of LWRWIMS for this type of calculation are made. (author)

  11. Transport fuel demand responses to fuel price and income projections : Comparison of integrated assessment models

    NARCIS (Netherlands)

    Edelenbosch, O. Y.; van Vuuren, Detlef; Bertram, C.; Carrara, S.; Emmerling, J.; Daly, H.; Kitous, A.; McCollum, D. L.; Saadi Failali, N.

    Income and fuel price pathways are key determinants in projections of the energy system in integrated assessment models. In recent years, more details have been added to the transport sector representation in these models. To better understand the model dynamics, this manuscript analyses transport

  12. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  13. A High Integrity Can Design for Degraded Nuclear Fuel

    International Nuclear Information System (INIS)

    Holmes, P.A.

    1999-01-01

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

  14. Evaluation of integrally finned cladding for LMFBR fuel pins

    International Nuclear Information System (INIS)

    Cantley, D.A.; Sutherland, W.H.

    1975-01-01

    An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

  15. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  16. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

    In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

  17. Optimization method of rod-type burnable poisons for nuclear designs of HTGRs

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu

    1994-01-01

    In block-type HTGRs, control rod insertion depths into cores had to be maintained as small as possible at full power operations, to avoid a fuel temperature rise. Thus, specifications (poison atom density (N BP ) and radius (r)) of rod-type burnable poisons (BPs) had to be optimized so that the effective multiplication factor (k eff ) would be constant at a minimum value throughout a planned burnup period. However, the optimization had been a time-consuming work until now since survey calculations had to be done for most possible combinations of N BP and r. To solve this problem, I have found a optimization method consisting of two steps. In the first step, approximation formulas describing a time-dependent relation among effective absorption cross sections (Σ aBP ), N BP and r are used to select promising combinations of N BP and r beforehand. In the second step, the best combination of N BP and r is determined by a comparison between Σ aBP of each promising combination and expected one. The number of survey calculations was reduced to about 1/10 by the optimization method. The change in k eff for 600 burnup days was reduced to 2%Δk by the method. Hence, it was made possible to operate reactors practically without inserting the control rods into cores. (author)

  18. Study of burnable poisons and gadolinium qualification in light water reactors

    International Nuclear Information System (INIS)

    Nasr, Mohamed.

    1981-09-01

    The aim of this work is to develop a calculation procedure for analyzing light water moderated reactors utilizing gadolinium as a burnable poison. The main points of this work can be summarized as follows: the available cross section data of gadolinium were analysed and corrected whenever it was necessary. The processes which include required precautions for obtaining multigroup cross sections were defined; an exhaustive study of the assumptions used in multicell calculation methods allowed the definition of option to be used for obtaining good results without excessive calculation cost. This study was followed by the interpretation of experimental results; when gadolinium is used in grain structure, a problem of double heterogeneity is encountered. A new calculation method was developed for such situations. Its validity was confirmed by a comparison with the Monte Carlo method; the problems encountered in performing a study of burn up of fuel elements containing gadolinium were analysed and the necessary precautions were established. The effect of the initial charge and geometrical form of the gadolinium and the behavior of lattices during the burn up were examined [fr

  19. IFBA credit in the Shearon Harris fuel racks with Vantage 5 fuel

    International Nuclear Information System (INIS)

    Boyd, W.A.; Schmidt, R.F.; Erwin, R.D.

    1989-01-01

    At the Shearon Harris nuclear plant, fuel management strategies are being considered which will result in feed fuel enrichments approaching 5.0 w/o U-235. These types of enrichments require a new criticality analysis to raise the existing fuel rack enrichment limit. It is receiving Westinghouse Vantage 5 fuel with integral fuel burnable absorber (IFBA) rods providing the depletable neutron absorber. An analysis was performed on the fuel racks which demonstrates that fuel enriched up to 5.0 w/o U-235 can be stored by taking credit for the IFBA rods present in the high enriched fuel assemblies. This is done by calculating the maximum Vantage 5 fuel assembly reactivity that can be placed in the fuel racks and meet the criticality K-eff limit. A methodology is also developed which conservatively calculates the minimum number of IFBA rods needed per assembly to meet the fuel rack storage limits. This eliminates the need for core designers to determine assembly K-inf terms for every different enrichment/IFBA combination

  20. A Study of Integrity Evaluation System for Spent Fuel and Selection of the Representative Spent Fuel

    International Nuclear Information System (INIS)

    Kim, J. G.; Lee, S. K.; Lim, C. J.; Kim, J. K.; Lee, S. J.

    2014-01-01

    Spent fuel (SF) integrity evaluation is a regulatory requirement that is described in 10 CFR 71(transportation) and 10 CFR 72(storage) of the U. S. NRC licensing requirement. NRC regulation states that retrievability of SF after storage should be ensured and SF integrity under the normal condition must be guaranteed during transportation and handling process that is entailed before/during/after the interim storage. And SF integrity evaluation under the hypothetical accident condition is a core technology element for an assessment of critical, shielding, and containment. In this paper, SF integrity evaluation system which is suitable for domestic situation is suggested, and necessity of representative SF selection and its method is described. The ultimate goal of the SF integrity evaluation is to evaluate a safety margin in case of transportation/ handling/storage of SFs. It means that retrievability of SF after storage should be assured and SF integrity must be guaranteed at normal condition in the process of transportation/handling accompanied before/during/after interim storage. In Korea, SF integrity evaluation system is not established up to date. Especially, representative SF selection technology that is essential to SF integrity evaluation has not been fulfilled. To overcome this situation effectively, the methodology and technology of an overseas agency need to be benchmarked. In this paper, an overseas SF integrity evaluation system is analyzed, and an evaluation system suitable for domestic situation is suggested. Also, necessity of representative SF selection and its method is described

  1. Integration of direct carbon and hydrogen fuel cells for highly efficient power generation from hydrocarbon fuels

    Energy Technology Data Exchange (ETDEWEB)

    Muradov, Nazim; Choi, Pyoungho; Smith, Franklyn; Bokerman, Gary [Florida Solar Energy Center, University of Central Florida, 1679 Clearlake Road, Cocoa, FL 32922-5703 (United States)

    2010-02-15

    In view of impending depletion of hydrocarbon fuel resources and their negative environmental impact, it is imperative to significantly increase the energy conversion efficiency of hydrocarbon-based power generation systems. The combination of a hydrocarbon decomposition reactor with a direct carbon and hydrogen fuel cells (FC) as a means for a significant increase in chemical-to-electrical energy conversion efficiency is discussed in this paper. The data on development and operation of a thermocatalytic hydrocarbon decomposition reactor and its coupling with a proton exchange membrane FC are presented. The analysis of the integrated power generating system including a hydrocarbon decomposition reactor, direct carbon and hydrogen FC using natural gas and propane as fuels is conducted. It was estimated that overall chemical-to-electrical energy conversion efficiency of the integrated system varied in the range of 49.4-82.5%, depending on the type of fuel and FC used, and CO{sub 2} emission per kW{sub el}h produced is less than half of that from conventional power generation sources. (author)

  2. An integrated expert system for optimum in core fuel management

    International Nuclear Information System (INIS)

    Abd Elmoatty, Mona S.; Nagy, M.S.; Aly, Mohamed N.; Shaat, M.K.

    2011-01-01

    Highlights: → An integrated expert system constructed for optimum in core fuel management. → Brief discussion of the ESOIFM Package modules, inputs and outputs. → Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). → The Package verification showed good agreement. - Abstract: An integrated expert system called Efficient and Safe Optimum In-core Fuel Management (ESOIFM Package) has been constructed to achieve an optimum in core fuel management and automate the process of data analysis. The Package combines the constructed mathematical models with the adopted artificial intelligence techniques. The paper gives a brief discussion of the ESOIFM Package modules, inputs and outputs. The Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). Moreover, the data of DNRR have been used as a case study for testing and evaluation of ESOIFM Package. This paper shows the comparison between the ESOIFM Package burn-up results, the DNRR experimental burn-up data, and other DNRR Codes burn-up results. The results showed good agreement.

  3. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  4. Enhanced canopy fuel mapping by integrating lidar data

    Science.gov (United States)

    Peterson, Birgit E.; Nelson, Kurtis J.

    2016-10-03

    BackgroundThe Wildfire Sciences Team at the U.S. Geological Survey’s Earth Resources Observation and Science Center produces vegetation type, vegetation structure, and fuel products for the United States, primarily through the Landscape Fire and Resource Management Planning Tools (LANDFIRE) program. LANDFIRE products are used across disciplines for a variety of applications. The LANDFIRE data retain their currency and relevancy through periodic updating or remapping. These updating and remapping efforts provide opportunities to improve the LANDFIRE product suite by incorporating data from other sources. Light detection and ranging (lidar) is uniquely suitable for gathering information on vegetation structure and spatial arrangement because it can collect data in three dimensions. The Wildfire Sciences Team has several completed and ongoing studies focused on integrating lidar into vegetation and fuels mapping.

  5. Control component structure and its removal from fuel assembly

    International Nuclear Information System (INIS)

    Edwards, G.T.; Schluderberg, D.C.

    1982-01-01

    This invention provides methods and apparatus for securing and removing burnable poison rods to the spider in a fuel assembly. A pin is secured to one of the transverse ends of a burnable poison rod. The pin is seated in a bore that is formed in the spider arm appropriate to the rod under consideration. The burnable poison rod is separated from the spider arm by applying a force in a direction that is coincident with the longitudinal axis of the rod and its associated pin. The force is of sufficient magnitude to press the pin out of the spider arm

  6. A Neutronic Feasibility Study of an OPR-1000 Core Design with Boron-bearing Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Park, Sang Yoon; Lee, Chung Chan; Yang, Yong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Westinghouse plants, boron is mainly used as a form of the integral fuel burnable absorber (IFBA) with a thin coating of zirconium diboride (ZrB{sub 2}) or wet annular burnable absorber (WABA) with a hollow Al{sub 2}O{sub 3}+B{sub 4}C pellet. In OPR-1000, on the other hand, gadolinia is currently employed as a form of an admixture which consists of Gd{sub 2}O{sub 3} of 6∼8 w/o and UO{sub 2} of natural uranium. Recently, boron-bearing UO{sub 2} fuel (BBF) with the high density of greater than 94%TD has been developed by using a low temperature sintering technique. In this paper, the feasibility of replacing conventional gadolinia-bearing UO{sub 2} fuel (GBF) in OPR-1000 with newly developed boron-bearing fuel is evaluated. Neutronic feasibility study to utilize the BBF in OPR-1000 core has been performed. The results show that the OPR-1000 core design with the BBF is feasible and promising in neutronic aspects. Therefore, the use of the BBF in OPR-1000 can reduce the dependency on the rare material such as gadolinium. However, the burnout of the {sup 10}B isotope results in helium gas, so fuel performance related study with respect to helium generation is needed.

  7. Purifier-integrated methanol reformer for fuel cell vehicles

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

    Pitt, L.

    2006-01-01

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

  9. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  10. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    International Nuclear Information System (INIS)

    Liang, C.; Ji, W.

    2013-01-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  11. Integrated multi-scale modelling and simulation of nuclear fuels

    International Nuclear Information System (INIS)

    Valot, C.; Bertolus, M.; Masson, R.; Malerba, L.; Rachid, J.; Besmann, T.; Phillpot, S.; Stan, M.

    2015-01-01

    This chapter aims at discussing the objectives, implementation and integration of multi-scale modelling approaches applied to nuclear fuel materials. We will first show why the multi-scale modelling approach is required, due to the nature of the materials and by the phenomena involved under irradiation. We will then present the multiple facets of multi-scale modelling approach, while giving some recommendations with regard to its application. We will also show that multi-scale modelling must be coupled with appropriate multi-scale experiments and characterisation. Finally, we will demonstrate how multi-scale modelling can contribute to solving technology issues. (authors)

  12. 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

  13. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  14. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  15. Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

    International Nuclear Information System (INIS)

    Penndorf, K.

    1976-04-01

    Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

  16. Integral logistics of the nuclear fuel Factory Juzbado

    International Nuclear Information System (INIS)

    Perez, P.

    2015-01-01

    The Logistic considers the complete process since the determination of possible demand, production planning, materials procurement, production control and delivery of final products to customer. This complete process is managed in all the scope under the same department called Planning and Logistic. This integration, some times really complex, has allowed to Enusa factory control all the key aspects that allow its running completely, considering the synergy's and important advantages to solve different problems. This article describes how we work of the main areas of procurement, production planning and control, fuel delivery and project planning of improvements on equipment's and factory systems, with an integrated management of all of them under the same direction. (Author)

  17. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  18. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    International Nuclear Information System (INIS)

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

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

  20. Performance evaluation of integrated fuel processor for residential PEMFCs application

    International Nuclear Information System (INIS)

    Yu Taek Seo; Dong Joo Seo; Young-Seog Seo; Hyun-Seog Roh; Wang Lai Yoon; Jin Hyeok Jeong

    2006-01-01

    KIER has been developing the natural gas fuel processor to produce hydrogen rich gas for residential PEMFCs system. To realize a compact and high efficiency, the unit processes of steam reforming, water gas shift, and preferential oxidation are chemically and physically integrated in a package. Current fuel processor designed for 1 kW class PEMFCs shows thermal efficiency of 78% as a HHV basis with methane conversion of 90% at rated load operation. CO concentration below 10 ppm in the produced gas is achieved with preferential oxidation unit using Pt and Ru based catalyst under the condition of [O 2 ]/[CO]=2.0. The partial load operation have been carried out to test the performance of fuel processor from 40% to 80% load, showing stable methane conversion and CO concentration below 10 ppm. The durability test for the daily start-stop and 8 hr operation procedure is under investigation and shows no deterioration of its performance after 40 start-stop cycles. (authors)

  1. Manufacturing method of fuel assembly and channel box for the fuel assembly

    International Nuclear Information System (INIS)

    Fujieda, Tadashi; Inagaki, Masatoshi; Takase, Iwao; Nishino, Yoshitaka; Yamashita, Jun-ichi; Yamanaka, Akihiro; Ito, Ken-ichi; Nakajima, Junjiro; Seto, Takehiro.

    1998-01-01

    An MOX fuel assembly to be used for a BWR type reactor comprises a channel box, a great number of fuel rod bundles and a water rod. BP members incorporated with a burnable neutron absorbing poison (BP) are buried in the vicinity of corners of four sides of the channel box in the longitudinal direction. The channel box is formed by fitting the BP members in concaves formed in the longitudinal direction of zircaloy plates, laminating other zircaloy plates and welding the seams. Then, hot rolling, cold rolling and annealing are conducted to form them into a single plate. Integrated two single plates after bending treatment are abutted and welded, and heat-treatment is applied to complete the channel box. With such a constitution, since the BP member is not brought into contact with reactor water directly, crevice corrosion or galvanic corrosion can be prevented. (I.N.)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  3. Integrating repositories with fuel cycles: The airport authority model

    International Nuclear Information System (INIS)

    Forsberg, C.

    2012-01-01

    The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuel fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members

  4. Integrating repositories with fuel cycles: The airport authority model

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139-4307 (United States)

    2012-07-01

    The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuel fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members

  5. MCNP apply in calculating reactor critical coefficient Keff under the changing of the burnable poison rod

    International Nuclear Information System (INIS)

    Wang Xinghua; Zhou Sichun; Zhang Qingxian; Zhao Feng; Liu Jun; Zhu Jian

    2013-01-01

    Taking Qinshan nuclear power plant as an example, in this paper, Monte Carlo method was used in the MCNP procedures for the establishment of nuclear power station simulation model, construct the reactor pressure vessel and vessel core component composition and arrangement, KCODE card was used to calculate the effect of the number and the location of burnable poison control rod factor K eff by the boron acid. The calculation results show that, with the increasing in the number of burnable poison control rod value-added factor K eff shown a downward trend, and with the burnable poison control rod from the dense to sparse, which K eff will be decreasing slowly. This condition is consistent with the theoretical. (authors)

  6. Integral approach to innovative fuel and material investigations in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.

    2009-01-01

    Integral approach used for fuel and material investigations in the Halden reactor can be used in support of qualification and certification of fuel to be introduced in commercial NPPs. This approach has been partly used for WWER fuel investigation in the Halden Reactor in a series of irradiation tests. In-pile fuel performance tests with reliable measurements provided by Halden instrumentation under different conditions can be used for validation of the WWER fuel behaviour models and verification of fuel performance codes. These models and codes can be used for qualification of innovative fuel behaviour under extended conditions

  7. Reactivity and neutron flux measurements in IPEN/MB-01 reactor with B4C burnable poison

    International Nuclear Information System (INIS)

    Fer, Nelson Custodio; Moreira, Joao Manoel Losada

    2000-01-01

    Burnable poison rods, made of B 4 C- Al 2 O 3 pellets with 5.01 mg/cm 3 10 B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutron flux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

  8. Rare earths as burnable poison for extended cycles control in electricity generation reactors; Etude des terres rares en tant que poison consommable pour le controle des cycles allonges pour les reacteurs electrogenes

    Energy Technology Data Exchange (ETDEWEB)

    Asou, M

    1995-05-12

    The search of an optimization of the French electronuclear network operations leads to a necessary optimization of the core performances. All the economic studies performed by the utilities had shown that there is a real gain to minimize shut down periods for refueling. So, increasing the cycle length from 12 to 18 months will present a gain of shut down for a three years operation period. The theoretical burnable absorber will be a fuel admixed material bringing the required initial negative reactivity with a burn-up kinetic well suited to the fuel and allowing the lowest residual penalty as possible. The residual penalty us defined in this case by the non complete burn up of the poison, by the low of fissile material and by the accumulate of residual isotopes or nuclides. Because of the well known use of gadolinium as burnable absorber for BWR`s and PWR`s operations, the search for the best compromise to optimize all the above stress is pointed towards the rare earths. In the nuclides family, considering criteria such as cross sections, natural abundance and availability only five nuclides can play the role as burnable absorbers, namely: gadolinium, samarium, dysprosium, europium and erbium. The study presented here will show that only gadolinium and erbium will be considered to control the reactivity of the PWR`s. (author). 58 refs., 65 figs., 47 tabs.

  9. Hydrogen, fuel cells and renewable energy integration in islands

    International Nuclear Information System (INIS)

    Bauen, A.; Hart, D.; Foradini, F.; Hart, D.

    2002-01-01

    Remote areas such as islands rely on costly and highly polluting diesel and heavy fuel oil for their electricity supply. This paper explored the opportunities for exploiting economically and environmentally viable renewable energy sources, in particular hydrogen storage, on such islands. In particular, this study focused on addressing the challenge of matching energy supply with demand and with technical issues regarding weak grids that are hindered with high steady state voltage levels and voltage fluctuations. The main technical characteristics of integrated renewable energy and hydrogen systems were determined by modelling a case study for the island of El Hierro (Canary Islands). The paper referred to the challenges regarding the technical and economic viability of such systems and their contribution to the economic development of remote communities. It was noted that energy storage plays an important role in addressing supply and demand issues by offering a way to control voltage and using surplus electricity at times of low load. Electrical energy can be stored in the form of potential or chemical energy. New decentralized generation technologies have also played a role in improving the energy efficiency of renewable energy sources. The feasibility of using hydrogen for energy storage was examined with particular reference to fuel-cell based energy supply in isolated island communities. 4 refs., 5 figs

  10. Nuclear analysis of the experimental VHTR fuel lattice

    International Nuclear Information System (INIS)

    Doi, Takeshi; Shindo, Ryuiti; Hirano, Mitsumasa; Takano, Makoto

    1984-11-01

    Nuclear properties of a fuel lattice in the experimental VHTR core were analyzed with DELIGHT-6 and SRAC codes. Analytical results by both codes were compared by using various calculational model. The nuclear parameters were analyzed, such as a multiplication factor of a fuel lattice and it's variation with burnup, a temperature effect on reactivity, an effect of double-heterogeniety in a resonance absorption calculation, a resonance integral of 238 U and a reactivity worth of burnable poison. From these analyses, following results were obtained. Firstly, on calculational models, 1) Effect of double-heterogeniety in the resonance absorption calculation for Mark-III fuel element, causing a decrease of about 5.5 barns in the resonance integral and an increase of about 2.6 %ΔK in the infinite multiplication factor, 2) The heterogeneous calculation with the collision probability method resulted in about 0.6 %ΔK higher the multiplication factor of fuel lattice than that with the point model, 3) The reactivity worth of burnable poison rod by a multi-region model is about 20 % less than that by a 2-region model at an initial state of burnup and it's variation with burnup are fairly different, Secondly, on comparison between the results by DELIGHT-6 and SRAC, 4) The nuclear parameters obtained with both codes agreed well, Lastly, on the improvement in DELIGHT-6, 5) Consideration of the neutron spectrum shielding effect in the resonance effective cross section calculation caused a decrease of about 2.4 %ΔK in the multiplication factor of fuel lattice, 6) The lattice multiplication factor increased about 0.5 %ΔK by introducing lambda-parameters for the non-resonant nuclie. (J.P.N.)

  11. Implementation of integrated safeguards at Nuclear Fuel Plant Pitesti, Romania

    International Nuclear Information System (INIS)

    Olaru, Vasilica; Ivana, Tiberiu; Epure, Gheorghe

    2010-01-01

    The nuclear activity in ROMANIA was for many years under Traditional Safeguards (TS) and has developed in good conditions this type of nuclear safeguards. Now, the opportunity exists to improve the performance and quality of the safeguards activity and increase the accountancy and control of nuclear material by passing to Integrated Safeguards (IS). The legal framework is the Law 100/2000 for ratification of the Protocol between Romania and International Atomic Energy Agency (IAEA), additional to the Agreement between the Socialist Republic of Romania Government and IAEA related to safeguards as part of the Treaty on the non-proliferation of nuclear weapons published in the Official Gazette no. 3/31 January 1970, and the Additional Protocol content published in the Official Gazette no. 295/ 29.06.2000. The first discussion about Integrated Safeguards (IS) between Nuclear Fuel Plant (NFP) representatives and IAEA inspectors was in June 2005. In Feb. 2007 an IAEA mission visited NFP and established the main steps for implementing the IS. There were visited the storages, technological flow, and was reviewed the disposal times for different nuclear materials, the applied chemical analysis, measuring methods, weighting method and elaborating procedure of the documents and lists. At that time the IAEA and NFP representatives established the main points for starting the IS at NFP: performing the Short Notice Random Inspections (SNRI); communication of the days established for SNRI for each year; communication of the estimated deliveries and shipments for first quarter and then for the rest of the year: daily mail box declaration (DD) with respect to the deposit time for several nuclear materials i.e. advance notification (AN) for each nuclear material transfer (shipments and receipts), others. At 01 June 2007 Romania has passed officially to Integrated Safeguards and NFP (WRMD) has taken all measures to implement this objective. (authors)

  12. 49 CFR 571.303 - Standard No. 303; Fuel system integrity of compressed natural gas vehicles.

    Science.gov (United States)

    2010-10-01

    ... compressed natural gas vehicles. 571.303 Section 571.303 Transportation Other Regulations Relating to... system integrity of compressed natural gas vehicles. S1. Scope. This standard specifies requirements for the integrity of motor vehicle fuel systems using compressed natural gas (CNG), including the CNG fuel...

  13. Thermodynamic analysis and optimization of IT-SOFC-based integrated coal gasification fuel cell power plants

    NARCIS (Netherlands)

    Romano, M.C.; Campanari, S.; Spallina, V.; Lozza, G.

    2011-01-01

    This work discusses the thermodynamic analysis of integrated gasification fuel cell plants, where a simple cycle gas turbine works in a hybrid cycle with a pressurized intermediate temperature–solid oxide fuel cell (SOFC), integrated with a coal gasification and syngas cleanup island and a bottoming

  14. Safety aspects of the using Gd as burnable poison in PWR's

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.

    1978-01-01

    The experience of BELGONUCLEAIRE in using Gd in LWR's has indicated the safety related advantages of this burnable poison. The successfully operation of the BR3 PWR power plant with 5% of Gd rods is presented and extrapolated to large PWR's. (authro)

  15. Effects of limestone petrography and calcite microstructure on OPC clinker raw meals burnability

    Science.gov (United States)

    Galimberti, Matteo; Marinoni, Nicoletta; Della Porta, Giovanna; Marchi, Maurizio; Dapiaggi, Monica

    2017-10-01

    Limestone represents the main raw material for ordinary Portland cement clinker production. In this study eight natural limestones from different geological environments were chosen to prepare raw meals for clinker manufacturing, aiming to define a parameter controlling the burnability. First, limestones were characterized by X-Ray Fluorescence, X-Ray Powder Diffraction and Optical Microscopy to assess their suitability for clinker production and their petrographic features. The average domains size and the microstrain of calcite were also determined by X-Ray Powder Diffraction line profile analysis. Then, each limestone was admixed with clay minerals to achieve the adequate chemical composition for clinker production. Raw meals were thermally threated at seven different temperatures, from 1000 to 1450 °C, to evaluate their behaviour on heating by ex situ X-Ray Powder Diffraction and to observe the final clinker morphology by Scanning Electron Microscopy. Results indicate the calcite microstrain is a reliable parameter to predict the burnability of the raw meals, in terms of calcium silicates growth and lime consumption. In particular, mixtures prepared starting from high-strained calcite exhibit a better burnability. Later, when the melt appears this correlation vanishes; however differences in the early burnability still reflect on the final clinker composition and texture.

  16. Substitution of the soluble boron reactivity control system of a pressurized water reactor by gadolinium burnable poisons

    International Nuclear Information System (INIS)

    Galperin, A.; Segev, M.; Radkowsky, A.

    1986-01-01

    The results are presented of a research project that is aimed at designing a gadolinium burnable poison (BP) system for complete reactivity control of a pressurized water reactor (PWR) core during the ''equilibrium'' cycle, resulting in the elimination of the soluble boron system, which represents a considerable saving in both capital and operating costs. A flat and strong negative moderator temperature coefficient is assured for a poison-free moderator. The design analysis of a core, heavily loaded with gadolinium BP rods, was based on a BGUCORE neutronic package and cluster model of a fuel assembly. The project objective was achieved by a novel lumped BP rod, designed as an annulus of gadolinium, clad by zirconium, and inserted into vacant guide thimbles of fresh-fuel assemblies. Specific combinations were found for the inner/outer radii of the poison ring, gadolinium densities, and number of rods per assembly, resulting in an almost flat criticality curve during the cycle. A reactivity swing of ≅1% ΔK can be easily controlled by an existing system of control rods. Comparison of the fuel cycle length of a gadolinium-controlled core with that of the reference, soluble, boron-controlled core indicated that there is no penalty due to residual poison at end of life. Unique guidelines for the fuel loading strategy were applied to find a practical fuel-shuffling scheme by which the design and operational constraints of a typical PWR core of current design were satisfied. Several problems should be solved for a practical implementation of the presented design relative to operational and safety requirements of the existing control rod system. Adequate movement of the regulating rods should be determined and shutdown margins of the safety rods should be ascertained. Final judgment of the feasibility of the concept may be made following the solution of these and other regulatory-related issues

  17. Fuels planning: science synthesis and integration; economic uses fact sheet 04: My Fuel Treatment Planner

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    In the face of rapidly changing public and political attitudes toward fire and fuel planning, one thing remains constant: the fuel planner is ultimately responsible for making decisions on the land. This fact sheet discusses the options for fuel treatments, and the need, development, and use of the MS Excel-based tool, My Fuel Treatment Planner.

  18. Forests at risk: integrating risk science into fuel management strategies.

    Science.gov (United States)

    Jonathan. Thompson

    2008-01-01

    The threat from wildland fire continues to grow across many regions of the Western United States. Drought, urbanization, and a buildup of fuels over the last century have contributed to increasing wildfire risk to property and highly valued natural resources. Fuel treatments, including thinning overly dense forests to reduce fuel and lower fire risk, have become a...

  19. Effect of fuel burnup on the mechanical safety coefficients

    International Nuclear Information System (INIS)

    Plyashkevich, V.Ju.; Sidorenko, V.D.; Shishkov, L.K.

    2001-01-01

    )In the paper the results of studies of changes in the process of campaign 'disturbances' of local heat flux and local fuel burnup, resulting from the 'mechanical' deviations in the composition and geometrical characteristics of fuel rods from the nominal are given. As example, the WWER-440 fuel assembly with burnable poisons used in the five-year fuel cycle is considered. The effect of deviations in fuel enrichment, fuel content, gadolinium content and geometrical size was studied (Authors)

  20. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  1. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    International Nuclear Information System (INIS)

    Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R and D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  2. INEL integrated spent nuclear fuel consolidation task team report

    International Nuclear Information System (INIS)

    Henry, R.N.; Clark, J.H.; Chipman, N.A.

    1994-01-01

    This document describes a draft plan and schedule to consolidate spent nuclear fuel (SNF) and special nuclear material (SNW) from aging storage facilities throughout the Idaho National Engineering Laboratory (INEL) to the Idaho Chemical Processing Plant (ICPP) in a safe, cost-effective, and expedient manner. A fully integrated and resource-loaded schedule was developed to achieve consolidation as soon as possible. All of the INEL SNF and SNM management task, projects, and related activities from fiscal year 1994 to the end of the consolidation period are logic-tied and integrated with each other. The schedule and plan are presented to initiate discussion of their implementation, which is expected to generate alternate concepts that can be evaluated using the methodology described in this report. Three perturbations to consolidating SNF as soon as possible are also explored. If the schedule is executed as proposed, the new and on-going consolidation activities will require about 6 years to complete and about $25.3M of additional funding. Reduced annual operating costs are expected to recover the additional investment in about 6.4 years. The total consolidation program as proposed will cost about $66.8M and require about 6 years to recover via reduced operating costs from retired SNF/SNM storage facilities. Detailed schedules and cost estimates for the Test Reactor Area Materials Test Reactor canal transfers are included as an example of the level of detail that is typical of the entire schedule (see Appendix D). The remaining work packages for each of the INEL SNF consolidation transfers are summarized in this document. Detailed cost and resource information is available upon request for any of the SNF consolidation transfers

  3. Design of a thermally integrated bioethanol-fueled solid oxide fuel cell system integrated with a distillation column

    Science.gov (United States)

    Jamsak, W.; Douglas, P. L.; Croiset, E.; Suwanwarangkul, R.; Laosiripojana, N.; Charojrochkul, S.; Assabumrungrat, S.

    Solid oxide fuel cell systems integrated with a distillation column (SOFC-DIS) have been investigated in this study. The MER (maximum energy recovery) network for SOFC-DIS system under the base conditions (C EtOH = 25%, EtOH recovery = 80%, V = 0.7 V, fuel utilization = 80%, T SOFC = 1200 K) yields Q Cmin = 73.4 and Q Hmin = 0 kW. To enhance the performance of SOFC-DIS, utilization of internal useful heat sources from within the system (e.g. condenser duty and hot water from the bottom of the distillation column) and a cathode recirculation have been considered in this study. The utilization of condenser duty for preheating the incoming bioethanol and cathode recirculation for SOFC-DIS system were chosen and implemented to the SOFC-DIS (CondBio-CathRec). Different MER designs were investigated. The obtained MER network of CondBio-CathRec configuration shows the lower minimum cold utility (Q Cmin) of 55.9 kW and total cost index than that of the base case. A heat exchanger loop and utility path were also investigated. It was found that eliminate the high temperature distillate heat exchanger can lower the total cost index. The recommended network is that the hot effluent gas is heat exchanged with the anode heat exchanger, the external reformer, the air heat exchanger, the distillate heat exchanger and the reboiler, respectively. The corresponding performances of this design are 40.8%, 54.3%, 0.221 W cm -2 for overall electrical efficiency, Combine Heat and Power (CHP) efficiency and power density, respectively. The effect of operating conditions on composite curves on the design of heat exchanger network was investigated. The obtained composite curves can be divided into two groups: the threshold case and the pinch case. It was found that the pinch case which T SOFC = 1173 K yields higher total cost index than the CondBio-CathRec at the base conditions. It was also found that the pinch case can become a threshold case by adjusting split fraction or operating at

  4. Fuel pin failure root causes and power distribution gradients in WWER cores

    International Nuclear Information System (INIS)

    Mikus, J.

    2008-01-01

    The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

  5. Integration of the military and civilian nuclear fuel cycles in Russia

    International Nuclear Information System (INIS)

    Bukharin, O.

    1994-01-01

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

  6. A comparison of integral block and tubular interacting fuel element concepts for low enrichment HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A

    1972-04-15

    The tubular interacting fuel element has to date been the favoured U.K. high temperature reactor design. Recent attempts to lower fuel costs and the progress of the Fort St. Vrain reactor has focussed attention on alternative designs, and in particular on the attractive design simplicity of the integral block concept. The aim of this investigation is to compare the merits of both concepts from fuel cycle cost and thermal performance viewpoints and to determine whether optimization of the integral block concept leads to changes in the current design values of (a) fuel density, (b) Nc/Nu, and/or (c) mean discharge irradiation within the framework of present design limits.

  7. Evaluation of mechanical integrity for helical coil hold-down spring of PLUS7TM fuel

    International Nuclear Information System (INIS)

    Choi, Ki Sung; Kim, Yong Hwan; Kwon, Jung Tack; Kim, Kyu Tae

    2004-01-01

    Nuclear fuel assembly is subject to hydraulic forces generated by primary coolant flow during reactor operation. These forces may be sufficient to overcome the fuel assembly weight thereby allowing the fuel assembly to lift off of its support. To provide a positive hold-down margin against the upward coolant flow forces, helical coil springs or leaf springs are installed in the fuel assemblies. An advanced fuel for Korean Standard Nuclear Power Plants (KSNP), PLUS7 fuel has developed to get the thermal margin increase, failure free and high seismic resistance, etc. And the new designed helical coil hold-down spring was introduced into PLUS7 fuel assembly. The purpose of this paper is to evaluate the mechanical integrity for the helical coil hold-down spring of PLUS7 fuel assembly

  8. A micro fuel reformer integrated with a combustor and a microchannel evaporator

    Science.gov (United States)

    Yoshida, Kazushi; Tanaka, Shuji; Hiraki, Hisashi; Esashi, Masayoshi

    2006-09-01

    This paper describes the development of a micro fuel reformer integrated with a combustor and an evaporator. Fuel reforming tests were performed by using a mixture of methanol and water as reforming fuel and hydrogen as combustion fuel. It was found that the design of the microchannel evaporator is critical to obtain larger hydrogen output. Hydrogen output and CO concentration were investigated by varying the input combustion power at different fuel feeding rates. 32.9 sccm of hydrogen, which is equivalent to 5.9 W in lower heating value, was produced, when input combustion power was 11 W.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  11. Pressurized solid oxide fuel cell integral air accumular containment

    Science.gov (United States)

    Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

    2004-02-10

    A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

  12. MEMS-based fuel cells with integrated catalytic fuel processor and method thereof

    Science.gov (United States)

    Jankowski, Alan F [Livermore, CA; Morse, Jeffrey D [Martinez, CA; Upadhye, Ravindra S [Pleasanton, CA; Havstad, Mark A [Davis, CA

    2011-08-09

    Described herein is a means to incorporate catalytic materials into the fuel flow field structures of MEMS-based fuel cells, which enable catalytic reforming of a hydrocarbon based fuel, such as methane, methanol, or butane. Methods of fabrication are also disclosed.

  13. Exergy analysis of an integrated fuel processor and fuel cell (FP-FC) system

    NARCIS (Netherlands)

    Delsman, E.R.; Uju, C.U.; Croon, de M.H.J.M.; Schouten, J.C.; Ptasinski, K.J.

    2006-01-01

    Fuel cells have great application potential as stationary power plants, as power sources in transportation, and as portable power generators for electronic devices. Most fuel cells currently being developed for use in vehicles and as portable power generators require hydrogen as a fuel. Chemical

  14. Fuels planning: science synthesis and integration; economic uses fact sheet 03: economic impacts of fuel treatments

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    With increased interest in reducing hazardous fuels in dry inland forests of the American West, agencies and the public will want to know the economic impacts of fuel reduction treatments. This fact sheet discusses the economic impact tool, a component of My Fuel Treatment Planner, for evaluating economic impacts.

  15. On-line fuel and control rod integrity management in BWRs

    International Nuclear Information System (INIS)

    Larsson, Irina; Sihver, Lembit

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-15

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

  18. Application of fuel cells with heat recovery for integrated utility systems

    Science.gov (United States)

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

    1975-01-01

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

  19. Neutron physical investigations on the use of burnable poisons and gray absorber rods in large pressurized water reactors

    International Nuclear Information System (INIS)

    Brosche, C.; Katinger, T.; Kollmar, W.; Thieme, K.; Wagner, M.R.

    1977-11-01

    Methods and results of neutron physics calculations are described using burnable poisons and gray absorber rods in large PWR's. Calculated and measured values are compared, the effort for programming has been guessed. (orig.) [de

  20. Economic prospects of the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

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

    1991-01-01

    The IFR fuel cycle based on pyroprocessing involves only few operational steps and the batch-oriented process equipment systems are compact. This results in major cost reductions in all of three areas of reprocessing, fabrication, and waste treatment. This document discusses the economic aspects of this fuel cycle

  1. Operating experience with Exxon nuclear advanced fuel assembly and fuel cycle designs in PWRs

    International Nuclear Information System (INIS)

    Skogen, F.B.; Killgore, M.R.; Holm, J.S.; Brown, C.A.

    1986-01-01

    Exxon Nuclear Company (ENC) has achieved a high standard of performance in its supply of fuel reloads for both BWRs and PWRs, while introducing substantial innovations aimed at realization of improved fuel cycle costs. The ENC experience with advanced design features such as the bi-metallic spacer, the dismountable upper tie plate, natural uranium axial blankets, optimized water-to-fuel designs, annular pellets, gadolinia burnable absorbers, and improved fuel management scenarios, is summarized

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  4. Life cycle assessment integrated with thermodynamic analysis of bio-fuel options for solid oxide fuel cells.

    Science.gov (United States)

    Lin, Jiefeng; Babbitt, Callie W; Trabold, Thomas A

    2013-01-01

    A methodology that integrates life cycle assessment (LCA) with thermodynamic analysis is developed and applied to evaluate the environmental impacts of producing biofuels from waste biomass, including biodiesel from waste cooking oil, ethanol from corn stover, and compressed natural gas from municipal solid wastes. Solid oxide fuel cell-based auxiliary power units using bio-fuel as the hydrogen precursor enable generation of auxiliary electricity for idling heavy-duty trucks. Thermodynamic analysis is applied to evaluate the fuel conversion efficiency and determine the amount of fuel feedstock needed to generate a unit of electrical power. These inputs feed into an LCA that compares energy consumption and greenhouse gas emissions of different fuel pathways. Results show that compressed natural gas from municipal solid wastes is an optimal bio-fuel option for SOFC-APU applications in New York State. However, this methodology can be regionalized within the U.S. or internationally to account for different fuel feedstock options. Copyright © 2012 Elsevier Ltd. All rights reserved.

  5. Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

    2003-01-01

    To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

  6. Influence of the resonance integral value on the fuel cycle characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Graziani, G; Trauwaert, E

    1972-04-24

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

  7. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions

  8. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

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

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs

  9. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

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

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. (author)

  10. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

  11. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  12. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  13. Integrated quality status and inventory tracking system for FFTF driver fuel pins

    International Nuclear Information System (INIS)

    Gottschalk, G.P.

    1979-11-01

    An integrated system for quality status and inventory tracking of Fast Flux Test Facility (FFTF) driver fuel pins has been developed. Automated fuel pin identification systems, a distributed computer network, and a data base are used to implement the tracking system

  14. Molten carbonate fuel cell integral matrix tape and bubble barrier

    International Nuclear Information System (INIS)

    Reiser, C.A.; Maricle, D.L.

    1983-01-01

    A molten carbonate fuel cell matrix material is described made up of a matrix tape portion and a bubble barrier portion. The matrix tape portion comprises particles inert to molten carbonate electrolyte, ceramic particles and a polymeric binder, the matrix tape being flexible, pliable and having rubber-like compliance at room temperature. The bubble barrier is a solid material having fine porosity preferably being bonded to the matrix tape. In operation in a fuel cell, the polymer binder burns off leaving the matrix and bubble barrier providing superior sealing, stability and performance properties to the fuel cell stack

  15. An Integrated Model for Identifying Linkages Between the Management of Fuel Treatments, Fire and Ecosystem Services

    Science.gov (United States)

    Bart, R. R.; Anderson, S.; Moritz, M.; Plantinga, A.; Tague, C.

    2015-12-01

    Vegetation fuel treatments (e.g. thinning, prescribed burning) are a frequent tool for managing fire-prone landscapes. However, predicting how fuel treatments may affect future wildfire risk and associated ecosystem services, such as forest water availability and streamflow, remains a challenge. This challenge is in part due to the large range of conditions under which fuel treatments may be implemented, as response is likely to vary with species type, rates of vegetation regrowth, meteorological conditions and physiographic properties of the treated site. It is also due to insufficient understanding of how social factors such as political pressure, public demands and economic constraints affect fuel management decisions. To examine the feedbacks between ecological and social dimensions of fuel treatments, we present an integrated model that links a biophysical model that simulates vegetation and hydrology (RHESSys), a fire spread model (WMFire) and an empirical fuel treatment model that accounts for agency decision-making. We use this model to investigate how management decisions affect landscape fuel loads, which in turn affect fire severity and ecosystem services, which feedback to management decisions on fuel treatments. We hypothesize that this latter effect will be driven by salience theory, which predicts that fuel treatments are more likely to occur following major wildfire events. The integrated model provides a flexible framework for answering novel questions about fuel treatments that span social and ecological domains, areas that have previously been treated separately.

  16. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

  17. Nuclear criticality safety: general. 5. Reactivity Effect of Burnable Absorbers in Burnup Credit for the CASTOR X/32S Storage and Transport Cask

    International Nuclear Information System (INIS)

    Rombough, Charles T.; Lancaster, Dale B.; Diersch, Rudolf; Spilker, Harry

    2001-01-01

    When considering burnup credit in the licensing of storage and transportation casks, a significant effect is whether or not the burned fuel was depleted with burnable absorbers present. This paper presents the results of detailed calculations to quantitatively determine the burnable absorber effect for the CASTOR X/32S transport cask, which assumes burnup of the fuel in the criticality analysis. A radial view of the CASTOR X/32S cask is shown in Fig. 1. This is the actual plot of the geometry as modeled in KENO V.a. Note that there are no water-filled flux traps and the assemblies are tightly packed. This reduces the overall dimensions of the cask for a given number of fuel assemblies. Reactivity is held down by borated aluminum plates between the fuel assemblies and by placing absorber rod modules (ARMs) in the guide tubes of selected assemblies. If burnup of the fuel is not considered and the initial enrichment is 5.0 wt% 235 U, then 28 of the 32 fuel assemblies must contain an ARM to maintain a k eff 3 ; 4. moderator temperature of 604 K; 5. cooling time of 9.5 yr; 6. specific power of 60 W/g of U metal; 7. conservative axial and radial burnup shape distribution; 8. Westinghouse BP material containing 12.5 wt% B 4 C. Using the model described earlier, calculations were performed with varying numbers of BP fingers inserted for different exposure times. The results are shown in Tables I and II. The 1 s statistical error in these results is σ equals ±0.05%. Note that the BP finger and exposure effects decrease with fuel burnup and the effect is smaller when the cask contains ARMs. Conservatively combining the results from Tables I and II and interpolating, we can equate fewer BP fingers with longer BP exposure time as shown in Table III. The Table III results were checked by running the actual cases (for example, 20 BP fingers for 24 GWd/tonne exposure) to verify that the k eff 's for the cask were always less than the base-case values. These results can also be

  18. Modification of Japanese first nuclear ship reactor for a regional energy supply system using gadolinia as a burnable poison

    International Nuclear Information System (INIS)

    Sato, Kotaro; Shimazu, Yoichiro; Narabayashi, Tadashi; Tsuji, Masashi

    2009-01-01

    In our laboratory, a small regional energy supply system which uses a small nuclear reactor has been studied for a long time. This system could supply not only heat but also electricity. Heat could be used for hot-water supply, a heating system of a house, melting snow and so on. In this point, this system seems to be useful for the places like northern part of Japan where it snows in winter. This reactor is based on Nuclear Ship Mutsu which was developed as the first nuclear ship of Japan about 40 years ago. It has several advantages for a small reactor. For example, its moderator temperature coefficient is always to be deeply negative because boric acid solution is not used in moderator and coolant. This can lead to a self-controlled operation without control rod maneuvering for load change. But some modifications have been performed in order to satisfy requirements such as (1) longer core life without refueling and reshuffling, (2) reactivity adjustment for load change without control rods or soluble boron, (3) simpler operations for load changes and (4) ultimate safety with sufficient passive capability. In our previous study, we confirmed the core based on Mutsu core had longer core life (about 10 years) using high uranium enrichment fuel (more than 5wt%) and current 17x17 fuel assemblies. We also confirmed excess reactivity during the cycle could be suppressed using combination of erbium oxide (Er 2 O 3 ) and gadolinium oxide (Gd 2 O 3 ) as burnable poisons. Er 2 O 3 has advantages such that criticality safety can be kept even if uranium enrichment is more than 5wt% and burnup characteristics of the core can be gradual. But at this time there are 2 problems to apply for the core using Er 2 O 3 in Japan. First problem is that more than 5wt% enrichment fuel is not yet accepted in Japan. Second problem is that there are no experiences of using Er 2 O 3 in commercial reactors in Japan. Considering these problems, we have to modify the design of the core, using

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  20. Irradiation and corrosion behaviour of cadmium aluminate, a burnable poison for light water reactors

    International Nuclear Information System (INIS)

    Hattenbach, K.; Ahlf, J.; Hilgendorff, W.; Zimmermann, H.U.

    1979-01-01

    In quest of a cadmium containing material for use as burnable poison cadmium aluminate seemed promising. Therefore irradiation and corrosion experiments on specimens of cadmium aluminate in a matrix of aluminia were performed. Irradiation at 575 K and fast fluences up to 10 25 m -2 showed the material to have good radiation resistance and low swelling rates. Cadmium pluminate was resistant to corrosion attack in demineralized water of 575K. (orig.) [de

  1. DEEP DESULFURIZATION OF DIESEL FUELS BY A NOVEL INTEGRATED APPROACH

    Energy Technology Data Exchange (ETDEWEB)

    Xiaoliang Ma; Michael Sprague; Lu Sun; Chunshan Song

    2002-10-01

    In order to reduce the sulfur level in liquid hydrocarbon fuels for environmental protection and fuel cell applications, deep desulfurization of a model diesel fuel and a real diesel fuel was conducted by our SARS (selective adsorption for removing sulfur) process using the adsorbent A-2. Effect of temperature on the desulfurization process was examined. Adsorption desulfurization at ambient temperature, 24 h{sup -1} of LHSV over A-2 is efficient to remove dibenzothiophene (DBT) in the model diesel fuel, but difficult to remove 4-methyldibenzothiophene (4-MDBT) and 4,6-dimethyl-dibenzothiophene (4,6-DMDBT). Adsorption desulfurization at 150 C over A-2 can efficiently remove DBT, 4-MDBT and 4,6-DMDBT in the model diesel fuel. The sulfur content in the model diesel fuel can be reduced to less than 1 ppmw at 150 C without using hydrogen gas. The adsorption capacity corresponding to the break-through point is 6.9 milligram of sulfur per gram of A-2 (mg-S/g-A-2), and the saturate capacity is 13.7 mg-S/g-A-2. Adsorption desulfurization of a commercial diesel fuel with a total sulfur level of 47 ppmw was also performed at ambient temperature and 24 h{sup -1} of LHSV over the adsorbent A-2. The results show that only part of the sulfur compounds existing in the low sulfur diesel can be removed by adsorption over A-2 at such operating conditions, because (1) the all sulfur compounds in the low sulfur diesel are the refractory sulfur compounds that have one or two alkyl groups at the 4- and/or 6-positions of DBT, which inhibit the approach of the sulfur atom to the adsorption site; (2) some compounds coexisting in the commercial low sulfur diesel probably inhibit the interaction between the sulfur compounds and the adsorbent. Further work in determining the optimum operating conditions and screening better adsorbent is desired.

  2. Recent advances in PWR fuel design and performance experience at ABB-CENF

    International Nuclear Information System (INIS)

    Corsetti, Lawrence V.

    2004-01-01

    Utilities in the United States continue to move towards longer cycles and higher burnups to improve fuel cycle economics. This has placed increased demands for improved burnable absorber concepts. Zircaloy-4 corrosion behavior remains a high burnup performance concern. Over the past several years there has also been increasing utility interest in fuel reliability improvements. The development and application of erbia as a burnable absorber mixed directly with urania fuel will be discussed. Debris resistant fuel assembly designs and operating experience are reviewed. Oxide thickness measurements showing the improved corrosion resistance of optimized, low-tin Zircaloy-4 are presented. (author)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  4. Integral power evaluation in fossil fuel power plants; Evaluacion energetica integral en unidades de centrales termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa I, Luis R; Sanchez H, Laura E; Rodriguez M, Jose H [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico); Nebradt G, Jesus [Unidad de Investigacion y Desarrollo de la Subdireccion de Generacion de la Comision Federal de Electricidad, (Mexico)

    2006-07-01

    In this occasion, a methodology is presented that carries out an integral energy evaluation of fossil fuel power plants units (FFPPU) with the purpose of determining the root of the significant decrements of power produced soon after the annual maintenance service. This proposal, besides identifying the origin of the energy efficiency problems, offers information about the contributions of each one of the involved equipment in the total decrement of the unit. With this methodology, the maintenance focuses in the equipment that contributes to the greater energy loss. This document presents such methodology along with its application in a real case, results and necessary remedial actions, demonstrating that its application offers bases for the investment in corrective measures. [Spanish] En esta ocasion se presenta una metodologia que efectua una evaluacion energetica integral de las unidades de centrales termoelectricas (UCT) con el fin de determinar la raiz de los decrementos de potencia significativos producidos luego del servicio anual de mantenimiento. Dicha propuesta, ademas de identificar el origen de los problemas de eficiencia energetica, brinda informacion acerca de las aportaciones de cada uno de los equipos involucrados al decremento total de la unidad. Con esta metodologia, el mantenimiento se enfoca a los equipos que contribuyen a la mayor perdida de potencia. Este documento exhibe tal metodologia junto con su aplicacion en un caso real, resultados y las acciones correctivas necesarias, demostrando que su aplicacion ofrece bases para una inversion futura en medidas correctivas.

  5. Integrity, behavior and proposal of CARA fuel irradiation with empty negative coefficient

    International Nuclear Information System (INIS)

    Marino, Armando C.; Brasnarof, Daniel O.; Demarco, Gustavo L.; Agueda, Horacio C.

    2007-01-01

    The main issues of the CARA fuel, CVN version, are its negative void reactivity coefficient and an extraction burnup of ∼20000 MWd/ton U. The analysis of the fuel rod behaviour, under the irradiation conditions of the Embalse, Atucha I and II NPPs, are the key to recognize their demanding under operation, to review the classic issues of the PHWR fuels and to prepare a programme of experimental irradiations in order to demonstrate the CARA concept, to assess the fuel integrity, to improve the performance and the enhancement of the safety margins. (author) [es

  6. Analysis of a fuel cell on-site integrated energy system for a residential complex

    Science.gov (United States)

    Simons, S. N.; Maag, W. L.

    1979-01-01

    The energy use and costs of the on-site integrated energy system (OS/IES) which provides electric power from an on-site power plant and recovers heat that would normally be rejected to the environment is compared to a conventional system purchasing electricity from a utility and a phosphoric acid fuel cell powered system. The analysis showed that for a 500-unit apartment complex a fuel OS/IES would be about 10% more energy conservative in terms of total coal consumption than a diesel OS/IES system or a conventional system. The fuel cell OS/IES capital costs could be 30 to 55% greater than the diesel OS/IES capital costs for the same life cycle costs. The life cycle cost of a fuel cell OS/IES would be lower than that for a conventional system as long as the cost of electricity is greater than $0.05 to $0.065/kWh. An analysis of several parametric combinations of fuel cell power plant and state-of-art energy recovery systems and annual fuel requirement calculations for four locations were made. It was shown that OS/IES component choices are a major factor in fuel consumption, with the least efficient system using 25% more fuel than the most efficient. Central air conditioning and heat pumps result in minimum fuel consumption while individual air conditioning units increase it, and in general the fuel cell of highest electrical efficiency has the lowest fuel consumption.

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  8. An integrated approach to selecting materials for fuel cladding in advanced high-temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rangacharyulu, C., E-mail: chary.r@usask.ca [Univ. of Saskatchewan, Saskatoon, SK (Canada); Guzonas, D.A.; Pencer, J.; Nava-Dominguez, A.; Leung, L.K.H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    An integrated approach has been developed for selection of fuel cladding materials for advanced high-temperature reactors. Reactor physics, thermalhydraulic and material analyses are being integrated in a systematic study comparing various candidate fuel-cladding alloys. The analyses established the axial and radial neutron fluxes, power distributions, axial and radial temperature distributions, rates of defect formation and helium production using AECL analytical toolsets and experimentally measured corrosion rates to optimize the material composition for fuel cladding. The project has just been initiated at University of Saskatchewan. Some preliminary results of the analyses are presented together with the path forward for the project. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. Drop analysis for structural integrity evaluation of KJRR fuel transport container

    International Nuclear Information System (INIS)

    Yang, Yun Young; Lim, Jong Min; Choi, Woo Seok; Lee, Ju Chan

    2016-01-01

    A fuel transport container for KiJang Research Reactor(KJRR) has been developed to transport fresh fuel assemblies and fission molly targets which are used for a research reactor built in Kijang. The KJRR fuel transport container is a type-A(F) container, which is defined in domestic and foreign regulations of a radioactive substance container. According to Nuclear Safety and Security Commission's notification, the container should meet the accident conditions defined in IAEA safety Standard Series, US NRC and etc. In this study, a structural integrity of the KJRR fuel transport container is evaluated by conducting computational analyses of 9-meter free drop and 1 meter puncture. It is confirmed that structural integrity of the KJRR fuel transport container can be maintained in the transportation accident condition. Hereafter, when the test model is produced, a safety test will be conducted and its result will be compared with the result of drop and puncture analyses.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  13. System Losses Study - FIT (Fuel-cycle Integration and Tradeoffs)

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert S. Cherry; Denia Djokic; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros

    2010-09-01

    This team aimed to understand the broad implications of changes of operating performance and parameters of a fuel cycle component on the entire system. In particular, this report documents the study of the impact of changing the loss of fission products into recycled fuel and the loss of actinides into waste. When the effort started in spring 2009, an over-simplified statement of the objective was “the number of nines” – how would the cost of separation, fuel fabrication, and waste management change as the number of nines of separation efficiency changed. The intent was to determine the optimum “losses” of TRU into waste for the single system that had been the focus of the Global Nuclear Energy Program (GNEP), namely sustained recycle in burner fast reactors, fed by transuranic (TRU) material recovered from used LWR UOX-51 fuel. That objective proved to be neither possible (insufficient details or attention to the former GNEP options, change in national waste management strategy from a Yucca Mountain focus) nor appropriate given the 2009-2010 change to a science-based program considering a wider range of options. Indeed, the definition of “losses” itself changed from the loss of TRU into waste to a generic definition that a “loss” is any material that ends up where it is undesired. All streams from either separation or fuel fabrication are products; fuel feed streams must lead to fuels with tolerable impurities and waste streams must meet waste acceptance criteria (WAC) for one or more disposal sites. And, these losses are linked in the sense that as the loss of TRU into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. The effort has provided a mechanism for connecting these three Campaigns at a technical level that had not previously occurred – asking smarter and smarter questions, sometimes answering them, discussing assumptions, identifying R&D needs, and gaining new insights. The FIT model has been a

  14. Transient performance simulation of aircraft engine integrated with fuel and control systems

    International Nuclear Information System (INIS)

    Wang, C.; Li, Y.G.; Yang, B.Y.

    2017-01-01

    Highlights: • A new performance simulation method for engine hydraulic fuel systems is introduced. • Time delay of engine performance due to fuel system model is noticeable but small. • The method provides details of fuel system behavior in engine transient processes. • The method could be used to support engine and fuel system designs. - Abstract: A new method for the simulation of gas turbine fuel systems based on an inter-component volume method has been developed. It is able to simulate the performance of each of the hydraulic components of a fuel system using physics-based models, which potentially offers more accurate results compared with those using transfer functions. A transient performance simulation system has been set up for gas turbine engines based on an inter-component volume (ICV) method. A proportional-integral (PI) control strategy is used for the simulation of engine controller. An integrated engine and its control and hydraulic fuel systems has been set up to investigate their coupling effect during engine transient processes. The developed simulation system has been applied to a model aero engine. The results show that the delay of the engine transient response due to the inclusion of the fuel system model is noticeable although relatively small. The developed method is generic and can be applied to any other gas turbines and their control and fuel systems.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Assessment of clad integrity of PHWR fuel pin following a postulated severe accident

    International Nuclear Information System (INIS)

    Dutta, B.K.; Kushwaha, H.S.; Venkat Raj, V.

    2000-01-01

    A mechanistic fuel performance analysis code FAIR has been developed. The code can analyse fuel pins with free standing as well as collapsible clad under normal, off-normal and accident conditions of reactors. The code FAIR is capable of analysing the effects of high burnup on fuel behaviour. The code incorporates finite element based thermo-mechanical module for computing transient temperature distribution and thermal-elastic-plastic stresses in the fuel pin. A number of high temperature thermo-physical and thermo-mechanical models also have been incorporated for analysing fuel pins subjected to severe accident scenario. The present paper describes salient features of code FAIR and assessment of clad integrity of PHWR fuel pins with different initial burnup subjected to severe accident scenario. (author)

  17. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  18. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Heat recovery subsystem and overall system integration of fuel cell on-site integrated energy systems

    Science.gov (United States)

    Mougin, L. J.

    1983-01-01

    The best HVAC (heating, ventilating and air conditioning) subsystem to interface with the Engelhard fuel cell system for application in commercial buildings was determined. To accomplish this objective, the effects of several system and site specific parameters on the economic feasibility of fuel cell/HVAC systems were investigated. An energy flow diagram of a fuel cell/HVAC system is shown. The fuel cell system provides electricity for an electric water chiller and for domestic electric needs. Supplemental electricity is purchased from the utility if needed. An excess of electricity generated by the fuel cell system can be sold to the utility. The fuel cell system also provides thermal energy which can be used for absorption cooling, space heating and domestic hot water. Thermal storage can be incorporated into the system. Thermal energy is also provided by an auxiliary boiler if needed to supplement the fuel cell system output. Fuel cell/HVAC systems were analyzed with the TRACE computer program.

  20. Spent fuel test. Climax data acquisition system integration report

    International Nuclear Information System (INIS)

    Nyholm, R.A.; Brough, W.G.; Rector, N.L.

    1982-06-01

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

  1. ClearFuels-Rentech Integrated Biorefinery Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, Joshua [Project Director

    2014-02-26

    The project Final Report describes the validation of the performance of the integration of two technologies that were proven individually on a pilot scale and were demonstrated as a pilot scale integrated biorefinery. The integrated technologies were a larger scale ClearFuels’ (CF) advanced flexible biomass to syngas thermochemical high efficiency hydrothermal reformer (HEHTR) technology with Rentech’s (RTK) existing synthetic gas to liquids (GTL) technology.

  2. Development of advanced spent fuel management process / criticality safety analysis for integrated mockup and metallized spent fuel storage

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Shin, Hee Sung; Shin, Young Joon; Bae, Kang Mok

    1999-02-01

    Benchmark calculation for SCALE4.3 CSAS6 module and burnup credit criticality analysis performed by CSAS6 module are described in this report. Calculation biases by the SCALE4.3 CSAS6 module for PWR spent fuel, metallized spent fuel and aqueous nuclear materials have been determined on the basis of the benchmark to be 0.011, 0.023 and 0.010, respectively. The maximum allowable multiplication factor for an integrated mockup and metallized spent fuel storage is conservatively determined to be 0.927. With the aid of this code system, K eff values as a function of metallization ratio for the integrated mockup have been calculated. The maximum values of K eff for normal and hypothetical accident conditions are 0.346 and 0.598, respectively, much less than the maximum allowable multiplication factor of 0.927. Besides, burnup credit criticality analysis has been performed for infinite arrays of square and hexagonal canisters containing metallized spent fuel rods with different canister wall thickness, canister surface-to-surface distance and water content. It is revealed that the effective multiplication factor for canister arrays as mentioned above is well below the subcritical limit regardless of external conditions when its wall thickness is over 9 mm. (Author). 37 refs., 27 tabs., 64 figs

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  4. Most advanced HTP fuel assembly design for EPR

    International Nuclear Information System (INIS)

    Francillon, Eric; Kiehlmann, Horst-Dieter

    2006-01-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  5. 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

  6. Integrated spent fuel storage and transportation system using NUHOMS

    International Nuclear Information System (INIS)

    Lehnert, R.; McConaghy, W.; Rosa, J.

    1990-01-01

    As utilities with nuclear power plants face increasing near term spent fuel store needs, various systems for dry storage such as the NUTECH Horizontal Modular Storage (NUHOMS) system are being implemented to augment existing spent fuel pool storage capacities. These decisions are based on a number of generic and utility specific considerations including both short term and long term economics. Since the US Department of Energy (DOE) is tasked by the Nuclear Waste Policy Act with the future responsibility of transporting spent fuel from commercial nuclear power plants to a Monitored Retrievable Storage (MRS) facility anchor a permanent geologic repository, the interfaces between the utilities at-reactor dry storage system and the DOE's away-from-reactor transportation system become important. This paper presents a study of the interfaces between the current at-reactor NUHOMS system and the future away-from-reactor DOE transportation system being developed under the Office of Civilian Radioactive Waste Management (OCRWM) program. 7 refs., 9 figs., 1 tab

  7. Method for distinguishing fuel pellets

    International Nuclear Information System (INIS)

    Sagami, Masaharu; Kurihara, Kunitoshi.

    1978-01-01

    Purpose: To distinguish correctly and efficiently the kind of fuel substance enclosed in a cladding tube. Method: Elements such as manganess 55, copper 65, vanadium 51, zinc 64, scandium 45 and the like, each having a large neutron absorption cross section and discharging gamma rays of inherent bright line spectra are applied to or mixed in fuel pellets of different kinds in uranium enrichment degree, plutonium concentration, burnable poison concentration or the like. These fuel rods are irradiated with neutron beams, and energy spectra of gamma rays discharged upon this occasion are observed to carry out distinguishing of fuel pellets. (Aizawa, K.)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. The optimum fuel and power distribution for a PWR burnup cycle

    International Nuclear Information System (INIS)

    Stillman, J.A.

    1989-01-01

    A method was developed to determine the optimum fuel and power distributions for a PWR burnup cycle. The backward diffusion calculation [1] and the Core-wise Green's Function [2] method were used for the core model which provided analytic derivatives for solving the nonlinear optimization problem using successive linear programming [3] methods. The solution algorithm consisted of a reverse depletion strategy which begins at the end of cycle and solves simultaneously for the optimal fuel and burnable absorber distributions while the core is depleted to the beginning of cycle. The resulting optimal solutions minimize the required fissile fuel inventory and burnable absorber loading for a PWR

  10. Review of the Effects of Normal Conditions of Transport on Spent Fuel Integrity in Transportation Casks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Junggoo; Yoo, Youngik; Lee, Seongki; Lim, Chaejoon [Korea Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2014-10-15

    Spent fuel(SF) storage capacity of each domestic nuclear power plant will reach a saturated state in the near future. Although there are several methods of SF disposal, interim storage is suggested as the most realistic and promising alternative. SF integrity evaluation is a regulatory requirement that is described in Part 71 of Code of Federal Regulations, Title 10 of the U..S. NRC licensing requirement. In this paper, the report is reviewed written by EPRI in US and it is helpful to a development of domestic SF integrity evaluation technology. EPRI report about integrity evaluation method on normal conditions of high burn-up spent fuel transport is reviewed. First, dynamic forces occurred in one-foot side drop are calculated. And deformation patterns and fuel rods responses by dynamic forces calculated from spent fuel and cask model are analyzed. It is shown that the damage of fuel rods is not occurred by the dynamic forces on normal conditions. Assembly distortion is not predicted, by virtue of the facts that the spacer grids do not experience significant permanent deformation. Axial forces, bending moments and pinch forces of fuel rods are calculated and compared with the results under the hypothetical accident conditions. No occurrence of transverse tearing mode that is the most serious damage mode in side drop case is predicted. Till now, in Korea, regulatory requirements related with structural integrity of spent fuel are not specified such as 10CFR71. To establish own regulation standards, producing and analyzing sufficient experimental data must be performed preferentially. Based on this, failure analysis and criteria establishment are necessary through modeling and analyzing of spent fuel.

  11. Environmental Assessment of Integrated Food and Cooking Fuel Production for a Village in Ghana

    DEFF Research Database (Denmark)

    Kamp, Andreas; Østergård, Hanne; Bolwig, Simon

    2016-01-01

    Small-scale farming in Ghana is typically associated with synthetic fertilizer dependence and soil degradation. The farmers often rely on wood fuel for cooking imported from outside the farmland, a practice that is associated with deforestation. Integration of food and energy production may...... be a holistic approach to solving these issues. We study four approaches to providing food and fuel for cooking in a small-scale farming community. Present practice (PP) of synthetic fertilizer based food production and provision of wood fuel from outside the farming area is compared to three modeled...

  12. The light-water-reactor version of the Uranus integral fuel-rod code

    International Nuclear Information System (INIS)

    Moreno, A.; Lassmann, K.

    1977-01-01

    The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

  13. Integrated data management system for radioactive waste and spent fuel in Korea

    International Nuclear Information System (INIS)

    Shin, Young Ho

    2001-03-01

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation, and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. So through the system, the five principles (independence, openness, clearance, efficiency and reliance) of safety regulation can be realized, and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted by providing reliable information, it can ensure an openness within the international nuclear community and efficiently support international agreements among contracting parties by operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible to holistic control and finally re-organize the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy is to integrate safe management and unit safe disposal. For this objectives, benchmark study was performed on similar data base system worldwide and data specification with major input/output data during the first phase of this project

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-31

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

  15. Integrated Solid Oxide Fuel Cell Power System Characteristics Prediction

    Directory of Open Access Journals (Sweden)

    Marian GAICEANU

    2009-07-01

    Full Text Available The main objective of this paper is to deduce the specific characteristics of the CHP 100kWe Solid Oxide Fuel Cell (SOFC Power System from the steady state experimental data. From the experimental data, the authors have been developed and validated the steady state mathematical model. From the control room the steady state experimental data of the SOFC power conditioning are available and using the developed steady state mathematical model, the authors have been obtained the characteristic curves of the system performed by Siemens-Westinghouse Power Corporation. As a methodology the backward and forward power flow analysis has been employed. The backward power flow makes possible to obtain the SOFC power system operating point at different load levels, resulting as the load characteristic. By knowing the fuel cell output characteristic, the forward power flow analysis is used to predict the power system efficiency in different operating points, to choose the adequate control decision in order to obtain the high efficiency operation of the SOFC power system at different load levels. The CHP 100kWe power system is located at Gas Turbine Technologies Company (a Siemens Subsidiary, TurboCare brand in Turin, Italy. The work was carried out through the Energia da Ossidi Solidi (EOS Project. The SOFC stack delivers constant power permanently in order to supply the electric and thermal power both to the TurboCare Company and to the national grid.

  16. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  17. Integration of a municipal solid waste gasification plant with solid oxide fuel cell and gas turbine

    DEFF Research Database (Denmark)

    Bellomare, Filippo; Rokni, Masoud

    2013-01-01

    An interesting source of producing energy with low pollutants emission and reduced environmental impact are the biomasses; particularly using Municipal Solid Waste (MSW) as fuel, can be a competitive solution not only to produce energy with negligible costs but also to decrease the storage...... in landfills. A Municipal Solid Waste Gasification Plant Integrated with Solid Oxide Fuel Cell (SOFC) and Gas Turbine (GT) has been studied and the plant is called IGSG (Integrated Gasification SOFC and GT). Gasification plant is fed by MSW to produce syngas by which the anode side of an SOFC is fed wherein...

  18. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  19. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    International Nuclear Information System (INIS)

    Young, Michael F.

    1999-01-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks

  20. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    Alencar, Donizete Anderson de

    2004-01-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  1. Operational method for demonstrating fuel loading integrity in a reactor having accessible 235U fuel

    International Nuclear Information System (INIS)

    Ward, D.R.

    1979-07-01

    The Health Physics Research Reactor is a small pulse reactor at the Oak Ridge National Laboratory. It is desirable for the operator to be able to demonstrate on a routine basis that all the fuel pieces are present in the reactor core. Accordingly, a technique has been devised wherein the control rod readings are recorded with the reactor at delayed critical and corrections are made to compensate for the effects of variations in reactor height above the floor, reactor power, core temperature, and the presence of any massive neutron reflectors. The operator then compares these readings with the values expected based on previous operating experience. If this routine operational check suggests that the core fuel loading might be deficient, a more rigorous follow-up may be made

  2. Integrated system of safety features for spent fuel interim storage

    International Nuclear Information System (INIS)

    Pantazi, Doina; Stanciu, Marcela; Mateescu, Silvia; Marin, Ion

    1999-01-01

    The design of the spent fuel interim storage facility (SFISF) must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility. To elaborate the safety documentation necessary for licensing, we were trying to chose the most appropriate approach related to safety features for SFISF, based on national and international regulations, standards and recommendations, as well as on the experience of other countries with similar facilities and finally, on our own experience in designing other nuclear objectives in Romania. The paper presents the issues that we consider important for the safety evaluation and are developed as a detailed diagram. The diagram contains in a logical succession the following issues: - fundamental principles of radioprotection; - fundamental safety principles of radioactive waste management; - safety objectives of SFISF; - safety criteria for SFISF; - safety requirements for SFISF; - siting criteria for SFISF; - siting requirements for SFISF. (authors)

  3. Portable Fuel Cell Battery Charger with Integrated Hydrogen Generator

    Energy Technology Data Exchange (ETDEWEB)

    Bossel, Ulf G. [CH-5452 Oberrohrdorf (Switzerland)

    1999-10-01

    A fully self-sufficient portable fuel cell battery charger has been designed, built, operated and is now prepared for commercialisation. The lightweight device is equipped with 24 circular polymer electrolyte cells of an innovative design. Each cell is a complete unit and can be tested prior to stacking. Hydrogen is admitted to the anode chamber from the centre of the cell. Air can reach the cathode by diffusion through a porous metal foam layer placed between cathode and separator plate. Soft seals surround the centre hole of the cells to separate hydrogen from air. Water vapour generated by the electrochemical conversion is released into the atmosphere via the porous metal foam on the cathode. All hydrogen fed to the dead-ended anode chamber is converted to electric power. The device is equipped with a chemical hydrogen generator. The fuel gas is formed by adding small amounts of water to a particular chemical compound which is contained in disposable cartridges. With one such cartridge enough hydrogen can be generated to operate CD-players, radios, recorders or portable computers for some hours, depending on the current drawn by the electronic device. The handy portable battery charger delivers about 10 W at 12 V DC. It is designed to be used in remote areas as autonomous power source for charging batteries used in radios, CD players, cellular telephones, radio transmitters, flash lights or model air planes. The power can also be used directly to provide light, sound or motion. Patents have been filed and partners are sought for commercialisation. (author) 4 figs.

  4. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  5. Implementation and operational experience of an integrated fuel information service at the BNFL THORP facility

    International Nuclear Information System (INIS)

    Robson, D.N.; Ramsden, P.N.

    1995-01-01

    BNFL's THORP Plant, which started active operations early in 1994, has contracts to reprocess 7000t(U) of fuel belonging to 33 customers in 9 countries in the UK, Europe and Japan during its first 10 years of operation. Contracts are in place or being negotiated, and further business sought after, with the expectation of extending THORP's operations well beyond the initial 10 years. An integrated data management service, for the fuel storage areas of BNFL's THORP Division, is being implemented to replace several, independent, systems. This Fuel Information Service (FIS) will bring the Nuclear Materials Accountancy and Safeguards Records together with the Operating Records into one database from which all Safeguards Reports will be made. BNFL's contractual and commercial data and technical data on the stored fuel, required to support the reprocessing business, will also be brought into the common database. FIS is the first stage in a project to integrate the Materials Management systems throughout the THORP nuclear recycling business including irradiated fuel receipt and storage, reprocessing and storage of products, mixed oxide fuel manufacture and the conditioning and storage of wastes

  6. Homogeneity of nuclear fuel containing burnable poison; Homogenost jedrskega goriva z gorljivim strupom

    Energy Technology Data Exchange (ETDEWEB)

    Loose, A; Susnik, D; Ilic, R [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1988-07-01

    In this work the results of the microstructural investigations of the influence of the Gd{sub 2}O{sub 3} contents and the sintering conditions on the formation of the homogeneous (U,Gd)O{sub 2} solid solution, are presented. For this purpose sintering conditions, microstructure and diffusivity in UO{sub 2} -Gd{sub 2}O{sub 3} , were studied. It was found that, with a suitable preparation of powders and longer sintering times in dry hydrogen atmosphere above 1700 deg C, a homogeneous (U,Gd)O{sub 2} solid solution can be obtained. (author)

  7. Integrity: A semi-mechanistic model for stress corrosion cracking of fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M; Hallgrimson, K; Macquarrie, J; Alavi, P [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Sato, S; Kinoshita, Y; Nishimura, T [Electric Power Development Co. Ltd., Tokyo (Japan)

    1997-08-01

    In this paper we describe the features, validation, and illustrative applications of a semi-mechanistic model, INTEGRITY, which calculates the probability of fuel defects due to stress corrosion cracking. The model expresses the defect probability in terms of fundamental parameters such as local stresses, local strains, and fission product concentration. The assessments of defect probability continue to reflect the influence of conventional parameters like ramped power, power-ramp, burnup and Canlub coating. In addition, the INTEGRITY model provides a mechanism to account for the impacts of additional factors involving detailed fuel design and reactor operation. Some examples of the latter include pellet density, pellet shape and size, sheath diameter and thickness, pellet/sheath clearance, coolant temperature and pressure, etc. The model has been fitted to a database of 554 power-ramp irradiations of CANDU fuel with and without Canlub. For this database the INTEGRITY model calculates 75 defects vs 75 actual defects. Similarly good agreements were noted in the different sub-groups of the data involving non-Canlub, thin-Canlub, and thick-Canlub fuel. Moreover, the shapes and the locations of the defect thresholds were consistent with all the above defects as well as with additional 14 ripple defects that were not in the above database. Two illustrative examples demonstrate how the defect thresholds are influenced by changes in the internal design of the fuel element and by extended burnup. (author). 19 refs, 7 figs.

  8. Integrity: A semi-mechanistic model for stress corrosion cracking of fuel

    International Nuclear Information System (INIS)

    Tayal, M.; Hallgrimson, K.; Macquarrie, J.; Alavi, P.; Sato, S.; Kinoshita, Y.; Nishimura, T.

    1997-01-01

    In this paper we describe the features, validation, and illustrative applications of a semi-mechanistic model, INTEGRITY, which calculates the probability of fuel defects due to stress corrosion cracking. The model expresses the defect probability in terms of fundamental parameters such as local stresses, local strains, and fission product concentration. The assessments of defect probability continue to reflect the influence of conventional parameters like ramped power, power-ramp, burnup and Canlub coating. In addition, the INTEGRITY model provides a mechanism to account for the impacts of additional factors involving detailed fuel design and reactor operation. Some examples of the latter include pellet density, pellet shape and size, sheath diameter and thickness, pellet/sheath clearance, coolant temperature and pressure, etc. The model has been fitted to a database of 554 power-ramp irradiations of CANDU fuel with and without Canlub. For this database the INTEGRITY model calculates 75 defects vs 75 actual defects. Similarly good agreements were noted in the different sub-groups of the data involving non-Canlub, thin-Canlub, and thick-Canlub fuel. Moreover, the shapes and the locations of the defect thresholds were consistent with all the above defects as well as with additional 14 ripple defects that were not in the above database. Two illustrative examples demonstrate how the defect thresholds are influenced by changes in the internal design of the fuel element and by extended burnup. (author). 19 refs, 7 figs

  9. A tomographic method for verification of the integrity of spent nuclear fuel

    International Nuclear Information System (INIS)

    Jacobsson, Staffan; Haakansson, Ane; Andersson, Camilla; Jansson, Peter; Baecklin, Anders

    1998-03-01

    A tomographic method for experimental investigation of the integrity of used LWR fuel has been developed. It is based on measurements of the gamma radiation from the fission products in the fuel rods. A reconstruction code of the algebraic type has been written. The potential of the technique has been examined in extensive simulations assuming a gamma-ray energy of either 0.66 MeV ( 137 Cs) or 1.27 MeV ( 154 Eu). The results of the simulations for BWR fuel indicate that single fuel rods or groups of rods replaced with water or fresh fuel can be reliably detected independent of their position in the fuel assembly using 137 Cs radiation. For PWR fuel the same result is obtained with the exception of the most central positions. Here the more penetrable radiation from 154 Eu must be used in order to allow a water channel to be distinguished from a fuel rod. The results of the simulations have been verified experimentally for a 8x8 BWR fuel assembly. Special equipment has been constructed and installed at the interim storage CLAB. The equipment allows the mapping of the radiation field around a fuel assembly with the aid of a germanium detector fitted with a collimator with a vertical slit. The intensities measured in 2520 detector positions were used as input for the reconstruction code used in the simulations. The results agreed very well with the simulations and revealed significantly a position containing a water channel in the central part of the assembly

  10. Integrated management platform of nuclear fuel storage and transportation based on RFID

    International Nuclear Information System (INIS)

    Song Yafeng; Ma Yanqin; Chen Liyu; Jiang Yong; Wu Jianlei; Yang Haibo; Zhang Haiyan

    2012-01-01

    This paper describes integrated system model to improve work efficiency and optimize control measures of nuclear fuel storage and transportation, RFID and information integration technology is introduced, traditional management processes are innovated in data acquisition and monitoring fields as well, system solutions and design model are given by emphasizing on the following key technologies: cascade protection of information system, security protocol of RFID information, algorithm of collision. (authors)

  11. Design, integration and demonstration of a 50 W JP8/kerosene fueled portable SOFC power generator

    Science.gov (United States)

    Cheekatamarla, Praveen K.; Finnerty, Caine M.; Robinson, Charles R.; Andrews, Stanley M.; Brodie, Jonathan A.; Lu, Y.; DeWald, Paul G.

    A man-portable solid oxide fuel cell (SOFC) system integrated with desulfurized JP8 partial oxidation (POX) reformer was demonstrated to supply a continuous power output of 50 W. This paper discusses some of the design paths chosen and challenges faced during the thermal integration of the stack and reformer in aiding the system startup and shutdown along with balance of plant and power management solutions. The package design, system capabilities, and test results of the prototype unit are presented.

  12. Hydrogen storage systems based on magnesium hydride: from laboratory tests to fuel cell integration

    Science.gov (United States)

    de Rango, P.; Marty, P.; Fruchart, D.

    2016-02-01

    The paper reviews the state of the art of hydrogen storage systems based on magnesium hydride, emphasizing the role of thermal management, whose effectiveness depends on the effective thermal conductivity of the hydride, but also depends of other limiting factors such as wall contact resistance and convective exchanges with the heat transfer fluid. For daily cycles, the use of phase change material to store the heat of reaction appears to be the most effective solution. The integration with fuel cells (1 kWe proton exchange membrane fuel cell and solid oxide fuel cell) highlights the dynamic behaviour of these systems, which is related to the thermodynamic properties of MgH2. This allows for "self-adaptive" systems that do not require control of the hydrogen flow rate at the inlet of the fuel cell.

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

    International Nuclear Information System (INIS)

    Johnson, T.R.

    1986-01-01

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

  14. Design considerations for a 10-kW integrated hydrogen-oxygen regenerative fuel cell system

    Science.gov (United States)

    Hoberecht, M. A.; Miller, T. B.; Rieker, L. L.; Gonzalez-Sanabria, O. D.

    1984-01-01

    Integration of an alkaline fuel cell subsystem with an alkaline electrolysis subsystem to form a regenerative fuel cell (RFC) system for low earth orbit (LEO) applications characterized by relatively high overall round trip electrical efficiency, long life, and high reliability is possible with present state of the art technology. A hypothetical 10 kW system computer modeled and studied based on data from ongoing contractual efforts in both the alkaline fuel cell and alkaline water electrolysis areas. The alkaline fuel cell technology is under development utilizing advanced cell components and standard Shuttle Orbiter system hardware. The alkaline electrolysis technology uses a static water vapor feed technique and scaled up cell hardware is developed. The computer aided study of the performance, operating, and design parameters of the hypothetical system is addressed.

  15. Exergy analysis of components of integrated wind energy / hydrogen / fuel cell

    International Nuclear Information System (INIS)

    Hernandez Galvez, G.; Pathiyamattom, J.S.; Sanchez Gamboa, S.

    2009-01-01

    Exergy analysis is made of three components of an integrated wind energy to hydrogen fuel cell: wind turbine, fuel cell (PEMFC) and electrolyzer (PEM). The methodology used to assess how affect the second law efficiency of the electrolyzer and the FC parameters as temperature and operating pressure and membrane thickness. It develop methods to evaluate the influence of changes in the air density and height of the tower on the second law efficiency of the turbine. This work represents a starting point for developing the global availability analysis of an integrated wind / hydrogen / fuel cells, which can be used as a tool to achieve the optimum design of the same. The use of this system contribute to protect the environment

  16. Current status of the transient integral fuel element performance code URANUS

    International Nuclear Information System (INIS)

    Preusser, T.; Lassmann, K.

    1983-01-01

    To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)

  17. Review of Current Criteria of Spent Fuel Rod Integrity during Dry Storage

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Sun Ki; Bang, Je Geon; Song, Kun Woo

    2006-01-01

    A PWR spent fuel has been stored in a wet storage pool in Korea. However, the amount of spent fuel is expected to exceed the capacity of a wet storage pool within 10∼15 years. From the early 1970's, a research on the PWR spent fuel dry storage started because the dry storage system has been economical compared with the wet storage system. The dry storage technology for Zircaloy-clad fuel was assessed and licensed in many countries such as USA, Canada, FRG and Switzerland. In the dry storage system, a clad temperature may be higher than in the wet storage system and can reach up to 400 .deg.. A higher clad temperature can cause cladding failures during the period of dry storage, and thus a dry storage related research has essentially dealt with the prevention of clad degradation. It is temperature and rod internal pressure that cause cladding failures through the mechanisms such as clad creep rupture, hydride re-orientation, and stress-corrosion cracking etc.. In this paper, the current licensing criteria are summarized for the PWR spent fuel dry storage system, especially on spent fuel rod integrity. And it is investigated that an application propriety of existing criteria to Korea spent fuel dry storage system

  18. Nuclear Fuel Cycle Analysis by Integrated AHP and TOPSIS Method Using an Equilibrium Model

    International Nuclear Information System (INIS)

    Yoon, S. R.; Choi, S. Y.; Koc, W. I.

    2015-01-01

    Determining whether to break away from domestic conflict surrounding nuclear power and step forward for public consensus can be identified by transparent policy making considering public acceptability. In this context, deriving the best suitable nuclear fuel cycle for Korea is the key task in current situation. Assessing nuclear fuel cycle is a multicriteria decision making problem dealing with multiple interconnected issues on efficiently using natural uranium resources, securing an environment friendliness to deal with waste, obtaining the public acceptance, ensuring peaceful uses of nuclear energy, maintaining economic competitiveness compared to other electricity sources, and assessing technical feasibility of advanced nuclear energy systems. This paper performed the integrated AHP and TOPSIS analysis on three nuclear fuel cycle options against 5 different criteria including U utilization, waste management, material attractiveness, economics, and technical feasibility. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once through cycle(PWR-OT), PWR-MOX cycle, Pyro- SFR cycle. These fuel cycles are most likely to be adopted in the foreseeable future. Analytic Hierarchy Process (AHP) and TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). The analyzed nuclear fuel cycle options include the once-through cycle, the PWR-MOX recycle, and the Pyro-SFR recycle

  19. Nuclear Fuel Cycle Analysis by Integrated AHP and TOPSIS Method Using an Equilibrium Model

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Choi, S. Y. [UNIST, Ulju (Korea, Republic of); Koc, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Determining whether to break away from domestic conflict surrounding nuclear power and step forward for public consensus can be identified by transparent policy making considering public acceptability. In this context, deriving the best suitable nuclear fuel cycle for Korea is the key task in current situation. Assessing nuclear fuel cycle is a multicriteria decision making problem dealing with multiple interconnected issues on efficiently using natural uranium resources, securing an environment friendliness to deal with waste, obtaining the public acceptance, ensuring peaceful uses of nuclear energy, maintaining economic competitiveness compared to other electricity sources, and assessing technical feasibility of advanced nuclear energy systems. This paper performed the integrated AHP and TOPSIS analysis on three nuclear fuel cycle options against 5 different criteria including U utilization, waste management, material attractiveness, economics, and technical feasibility. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once through cycle(PWR-OT), PWR-MOX cycle, Pyro- SFR cycle. These fuel cycles are most likely to be adopted in the foreseeable future. Analytic Hierarchy Process (AHP) and TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). The analyzed nuclear fuel cycle options include the once-through cycle, the PWR-MOX recycle, and the Pyro-SFR recycle.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

  2. Supercritical kinetic analysis in simplified system of fuel debris using integral kinetic model

    International Nuclear Information System (INIS)

    Tuya, Delgersaikhan; Obara, Toru

    2016-01-01

    Highlights: • Kinetic analysis in simplified weakly coupled fuel debris system was performed. • The integral kinetic model was used to simulate criticality accidents. • The fission power and released energy during simulated accident were obtained. • Coupling between debris regions and its effect on the fission power was obtained. - Abstract: Preliminary prompt supercritical kinetic analyses in a simplified coupled system of fuel debris designed to roughly resemble a melted core of a nuclear reactor were performed using an integral kinetic model. The integral kinetic model, which can describe region- and time-dependent fission rate in a coupled system of arbitrary geometry, was used because the fuel debris system is weakly coupled in terms of neutronics. The results revealed some important characteristics of coupled systems, such as the coupling between debris regions and the effect of the coupling on the fission rate and released energy in each debris region during the simulated criticality accident. In brief, this study showed that the integral kinetic model can be applied to supercritical kinetic analysis in fuel debris systems and also that it can be a useful tool for investigating the effect of the coupling on consequences of a supercritical accident.

  3. Thermodynamic investigation of an integrated gasification plant with solid oxide fuel cell and steam cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rokni, Masoud [Technical Univ. of Denmark, Lyngby (Denmark). Dept. of Mechanical Engineering, Thermal Energy System

    2012-07-01

    A gasification plant is integrated on the top of a solid oxide fuel cell (SOFC) cycle, while a steam turbine (ST) cycle is used as a bottoming cycle for the SOFC plant. The gasification plant was fueled by woodchips to produce biogas and the SOFC stacks were fired with biogas. The produced gas was rather clean for feeding to the SOFC stacks after a simple cleaning step. Because all the fuel cannot be burned in the SOFC stacks, a burner was used to combust the remaining fuel. The off-gases from the burner were then used to produce steam for the bottoming steam cycle in a heat recovery steam generator (HRSG). The steam cycle was modeled with a simple single pressure level. In addition, a hybrid recuperator was used to recover more energy from the HRSG and send it back to the SOFC cycle. Thus two different configurations were investigated to study the plants characteristic. Such system integration configurations are completely novel and have not been studied elsewhere. Plant efficiencies of 56% were achieved under normal operation which was considerably higher than the IGCC (Integrated Gasification Combined Cycle) in which a gasification plant is integrated with a gas turbine and a steam turbine. Furthermore, it is shown that under certain operating conditions, plant efficiency of about 62 is also possible to achieve. (orig.)

  4. Fuels planning: science synthesis and integration; environmental consequences fact sheet 05: prescriptions and fire effects

    Science.gov (United States)

    Melanie Miller

    2004-01-01

    Fuels planning: science synthesis and integration; environmental consequences fact sheet 5: prescriptions and fire effects. Miller, Melanie. 2004. Res. Note RMRS-RN-23-5-WWW. Fort Collins, CO: U.S. Department of Agriculture, Forest Service, Rocky Mountain Research Station. 2 p. While our understanding of the causes for variation in postfire effects is increasing, burn...

  5. Biorefineries to integrate fuel, energy and chemical production processes

    Directory of Open Access Journals (Sweden)

    Enrica Bargiacchi

    2007-12-01

    Full Text Available The world of renewable energies is in fast evolution and arouses political and public interests, especially as an opportunity to boost environmental sustainability by mitigation of greenhouse gas emissions. This work aims at examining the possibilities related to the development of biorefineries, where biomass conversion processes to produce biofuels, electricity and biochemicals are integrated. Particular interest is given to the production processes of biodiesel, bioethanol and biogas, for which present world situation, problems, and perspectives are drawn. Potential areas for agronomic and biotech researches are also discussed. Producing biomass for biorefinery processing will eventually lead to maximize yields, in the non food agriculture.

  6. Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and its implementation to the operating WWER-440 units

    International Nuclear Information System (INIS)

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, P.; Filimonov, P.

    2000-01-01

    Over the past few years in Russia the investigations aimed at the increase of the reliability, safety and efficiency of operation of the WWER-1000 reactors as well as of its competitiveness in the world market were carried out. In the frame of these investigations the four-year fuel cycle, based on advanced fuel assemblies with zirconium alloy spacer grids and guide tubes and with fuel pellet having a reduced diameter of the central hole (1,5 mm), has been developed. For the compensation of a part of excess reactivity, Gd 2 O 3 integrated burnable absorbers are used. CPS absorbing rods contain a combine absorber (B 4 C + Dy 2 O 3 *TiO 2 ). A part of depleted fuel is located on the core periphery. The algorithms controlling the reactor power and power distribution have been updated. For checking of the solutions adopted and for verification of code package developed at the RRC 'Kurchatov Institute' the wide-scale experimental operation of advanced FA and its individual components is carried out. (Authors)

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

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

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

  8. An integrated methodology to evaluate a spent nuclear fuel storage system

    International Nuclear Information System (INIS)

    Yoon, Jeong Hyoun

    2008-02-01

    This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

  9. Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Groudev, P.P.; Atanasova, B.P.

    2009-01-01

    This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies-the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR 'input model' for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP

  10. The effect of non-uniform fuel rod temperatures on effective resonance integrals

    International Nuclear Information System (INIS)

    Reichel, A.

    1961-06-01

    The effective resonance integral for heterogeneous lattices can be reduced to the effective resonance integral for an equivalent homogeneous system with a fairly well defined error depending on lump size and geometry. This report investigates the effect of a radial parabolic temperature variation in cylindrical lumps on the equivalent homogeneous effective resonance integral. Also determined is the equivalent uniform temperature to be taken in the usual formulae to allow for non-uniform fuel rod temperature. This effective temperature is found to be T eff. = T s + 4/9 (T c - T s ) where T s and T c are the surface and central temperatures of the lump. (author)

  11. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated

  12. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.

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

    International Nuclear Information System (INIS)

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

    1986-08-01

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

  14. CANDU fuel sheath integrity and oxide layer thickness determination by Eddy current technique

    International Nuclear Information System (INIS)

    Gheorghe, Gabriela; Man, Ion; Parvan, Marcel; Valeca, Serban

    2010-01-01

    This paper presents results concerning the integrity assessment of the fuel elements cladding and measurements of the oxide layer on sheaths, using the eddy current technique. Flaw detection using eddy current provides information about the integrity of fuel element sheath or presence of defects in the sheath produced by irradiation. The control equipment consists of a flaw detector with eddy currents, operable in the frequency range 10 Hz to 10 MHz, and a differential probe. The calibration of the flaw detector is done using artificial defects (longitudinal, transversal, external and internal notches, bored and unbored holes) obtained on Zircaloy-4 tubes identical to those out of which the sheath of the CANDU fuel element is manufactured (having a diameter of 13.08 mm and a wall thickness of 0.4 mm). When analyzing the behavior of the fuel elements' cladding facing the corrosion is important to know the thickness of the zirconium oxide layer. The calibration of the device measuring the thickness of the oxide layer is done using a Zircaloy-4 tube identical to that which the cladding of the CANDU fuel element is manufactured of, and calibration foils, as well. (authors)

  15. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  16. Integrated data management system for radioactive waste and spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Taek [Korea Power Engineering Co., Inc., Yongin (Korea, Republic of)

    2002-05-15

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. Through the system, the five principles(independence, openness, clearance, efficiency and reliance) of safety regulation can be realized and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted. By providing reliable information and openness within the international nuclear community can be ensured and efficient support of international agreements among contracting parties can be ensured. By operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible for holistic control and reorganization of the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy so as to integrate safe management and unit safe disposal. To meet this objectives, design of the database system structure and the study of input/output data validation and verification methodology was performed during the second phase of this project.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Saegusa, T [CRIEPI (Japan)

    2012-07-01

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

  19. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  20. Recent enhancements of the INSIGHT integrated in-core fuel management tool

    International Nuclear Information System (INIS)

    Akio, Yamamoto

    2001-01-01

    Recent enhancements of the INSIGHT system are described in this paper. The INSIGHT system is an integrated in-core fuel management tool for pressurized water reactors (PWRs) runs on UNIX workstations. The INSIGHT system provides various capabilities which contribute to reduce fuel cycle cost and workload of in-core fuel management tasks, i.e. core follow calculations, interactive loading pattern design, automated multicycle analysis and interface between detailed core calculation codes. To minimize engineers' workload, most of input data for analysis modules are automatically generated by the INSIGHT system through specification of calculation conditions in the graphic user interface. Recent enhancements of the INSIGHT system are mainly focused to improve efficiency of loading pattern optimization and flexibility of multicycle analyses. To increase optimization efficiency, a parallel calculation capability, various optimization theories, extension of heuristic rules, screening by neural networks and so on were incorporated in the loading pattern optimization module. The multicycle analyses module was rewritten to increase flexibility such as cycle dependent specification of loading pattern search methods and so on. The INSIGHT system is currently used by Japanese utilities not only for regular in-core fuel management tasks but also for strategic fuel management studies to reduce fuel cycle cost

  1. Criteria for the selection of graphites for HTR integral block fuel elements

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1980-01-01

    This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

  2. Dynamic modeling of gas turbines in integrated gasification fuel cell systems

    Science.gov (United States)

    Maclay, James Davenport

    2009-12-01

    Solid oxide fuel cell-gas turbine (SOFC-GT) hybrid systems for use in integrated gasification fuel cell (IGFC) systems operating on coal will stretch existing fossil fuel reserves, generate power with less environmental impact, while having a cost of electricity advantage over most competing technologies. However, the dynamic performance of a SOFC-GT in IGFC applications has not been previously studied in detail. Of particular importance is how the turbo-machinery will be designed, controlled and operated in such applications; this is the focus of the current work. Perturbation and dynamic response analyses using numerical SimulinkRTM models indicate that compressor surge is the predominant concern for safe dynamic turbo-machinery operation while shaft over-speed and excessive turbine inlet temperatures are secondary concerns. Fuel cell temperature gradients and anode-cathode differential pressures were found to be the greatest concerns for safe dynamic fuel cell operation. Two control strategies were compared, that of constant gas turbine shaft speed and constant fuel cell temperature, utilizing a variable speed gas turbine. Neither control strategy could eliminate all vulnerabilities during dynamic operation. Constant fuel cell temperature control ensures safe fuel cell operation, while constant speed control does not. However, compressor surge is more likely with constant fuel cell temperature control than with constant speed control. Design strategies that provide greater surge margin while utilizing constant fuel cell temperature control include increasing turbine design mass flow and decreasing turbine design inlet pressure, increasing compressor design pressure ratio and decreasing compressor design mass flow, decreasing plenum volume, decreasing shaft moment of inertia, decreasing fuel cell pressure drop, maintaining constant compressor inlet air temperature. However, these strategies in some cases incur an efficiency penalty. A broad comparison of cycles

  3. Thermodynamic Analysis of a Woodchips Gasification Integrated with Solid Oxide Fuel Cell and Stirling Engine

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2013-01-01

    Integrated gasification Solid Oxide Fuel Cell (SOFC) and Stirling engine for combined heat and power application is analysed. The target for electricity production is 120 kW. Woodchips are used as gasification feedstock to produce syngas which is utilized for feeding the SOFC stacks for electricity...... and suggested. Thermodynamic analysis shows that a thermal efficiency of 42.4% based on LHV (lower heating value) can be achieved. Different parameter studies are performed to analysis system behaviour under different conditions. The analysis show that increasing fuel mass flow from the design point results...

  4. Thermodynamic Investigation of an Integrated Gasification Plant with Solid Oxide Fuel Cell and Steam Cycles

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2012-01-01

    A gasification plant is integrated on the top of a solid oxide fuel cell (SOFC) cycle, while a steam turbine (ST) cycle is used as a bottoming cycle for the SOFC plant. The gasification plant was fueled by woodchips to produce biogas and the SOFC stacks were fired with biogas. The produced gas...... generator (HRSG). The steam cycle was modeled with a simple single pressure level. In addition, a hybrid recuperator was used to recover more energy from the HRSG and send it back to the SOFC cycle. Thus two different configurations were investigated to study the plants characteristic. Such system...

  5. Integration of A Solid Oxide Fuel Cell into A 10 MW Gas Turbine Power Plant

    Directory of Open Access Journals (Sweden)

    Denver F. Cheddie

    2010-04-01

    Full Text Available Power generation using gas turbine power plants operating on the Brayton cycle suffers from low efficiencies. In this work, a solid oxide fuel cell (SOFC is proposed for integration into a 10 MW gas turbine power plant, operating at 30% efficiency. The SOFC system utilizes four heat exchangers for heat recovery from both the turbine outlet and the fuel cell outlet to ensure a sufficiently high SOFC temperature. The power output of the hybrid plant is 37 MW at 66.2% efficiency. A thermo-economic model predicts a payback period of less than four years, based on future projected SOFC cost estimates.

  6. In-core fuel management code package validation for BWRs

    International Nuclear Information System (INIS)

    1995-12-01

    The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

  7. performance calculations of gadolinium oxide and boron nitride coated fuel

    International Nuclear Information System (INIS)

    Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

    1997-01-01

    A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

  8. The main conditions ensured problemless implementation of 235U high enriched fuel in Kozloduy NPP (Bulgaria) - WWER-1000 Units

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.; Minkova, K.; Michaylov, G.; Penev, P.; Gerchev, N.

    2009-01-01

    The collected water chemistry and radiochemistry data during the operation of the Kozloduy NPP Unit 5 for the period 2006-2009 (12-th, 13-th 14-th and 15-th fuel cycles) undoubtedly indicate for WWER-1000 Units (whose specific features are: Steam generators with austenitic stainless steel 08Cr18N10T tubing; Steam generators are with horizontal straight tubing and Fuel elements cladding material is Zr-1%Nb (Zr1Nb) alloy), that one realistic way for problemless implementation of 235 U high enriched fuel have been found. The main feature characteristics of this way are: Implementation of solid neutron burnable absorbers together with the dissolved in coolant neutron absorber - natural boric acid; Application of fuel cladding materials with enough corrosion resistance by the specific fuel cladding environment created by presence of SNB; Keeping of suitable coolant water chemistry which ensures low corrosion rates of core- and out-of-core- materials and limits in core (cladding) depositions and restricts out-of-core radioactivity buildup. The realization of this way in WWER-1000 Units in Kozloduy NPP was practically carried out through: 1) Implementation of Russian fuel assemblies TVSA which have as fuel cladding material E-110 alloy (Zr1Nb) with enough high corrosion resistance by presence of sub-cooled nucleate boiling (SNB) and use burnable absorber (Gd) integrated in the uranium-gadolinium (U-Gd 2 O 3 ) fuel (fuel rod with 5.0% Gd 2 O 3 ); 2) Development and implementation of water chemistry primary circuit guidelines, which require the relation between boric acid concentration and total alkalising agent concentrations to ensure coolant pH 300 = 7.0 - 7.2 values during the whole operation period. The above mentioned conditions by the passing of WWER-1000 Units in NPP Kozloduy to uranium fuel with 4.4% 235 U (TVSA fuel assemblies) practically ensured avoidance of the creation of the necessary conditions for AOA onset. The operational experience (2006-2009) of the

  9. Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

    International Nuclear Information System (INIS)

    Anghaie, S.; Smith, B.

    2004-01-01

    This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

  10. Wabash Valley Integrated Gasification Combined Cycle, Coal to Fischer Tropsch Jet Fuel Conversion Study

    Energy Technology Data Exchange (ETDEWEB)

    Shah, Jayesh [Lummus Technology Inc., Bloomfield, NJ (United States); Hess, Fernando [Lummus Technology Inc., Bloomfield, NJ (United States); Horzen, Wessel van [Lummus Technology Inc., Bloomfield, NJ (United States); Williams, Daniel [Lummus Technology Inc., Bloomfield, NJ (United States); Peevor, Andy [JM Davy, London (United Kingdom); Dyer, Andy [JM Davy, London (United Kingdom); Frankel, Louis [Canonsburgh, PA (United States)

    2016-06-01

    This reports examines the feasibility of converting the existing Wabash Integrated Gasification Combined Cycle (IGCC) plant into a liquid fuel facility, with the goal of maximizing jet fuel production. The fuels produced are required to be in compliance with Section 526 of the Energy Independence and Security Act of 2007 (EISA 2007 §526) lifecycle greenhouse gas (GHG) emissions requirements, so lifecycle GHG emissions from the fuel must be equal to or better than conventional fuels. Retrofitting an existing gasification facility reduces the technical risk and capital costs associated with a coal to liquids project, leading to a higher probability of implementation and more competitive liquid fuel prices. The existing combustion turbine will continue to operate on low cost natural gas and low carbon fuel gas from the gasification facility. The gasification technology utilized at Wabash is the E-Gas™ Technology and has been in commercial operation since 1995. In order to minimize capital costs, the study maximizes reuse of existing equipment with minimal modifications. Plant data and process models were used to develop process data for downstream units. Process modeling was utilized for the syngas conditioning, acid gas removal, CO2 compression and utility units. Syngas conversion to Fischer Tropsch (FT) liquids and upgrading of the liquids was modeled and designed by Johnson Matthey Davy Technologies (JM Davy). In order to maintain the GHG emission profile below that of conventional fuels, the CO2 from the process must be captured and exported for sequestration or enhanced oil recovery. In addition the power utilized for the plant’s auxiliary loads had to be supplied by a low carbon fuel source. Since the process produces a fuel gas with sufficient energy content to power the plant’s loads, this fuel gas was converted to hydrogen and exported to the existing gas turbine for low carbon power production. Utilizing low carbon fuel gas and

  11. The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

    Energy Technology Data Exchange (ETDEWEB)

    Labmann, K; Moreno, A

    1977-07-01

    The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

  12. Thermodynamic Analysis of an Integrated Gasification Solid Oxide Fuel Cell Plant with a Kalina Cycle

    DEFF Research Database (Denmark)

    Pierobon, Leonardo; Rokni, Masoud

    2015-01-01

    % is achieved; plant size and nominal power are selected based on the required cultivation area. SOFC heat recovery with SKC is compared to a Steam Cycle (SC). Although ammonia-water more accurately fits the temperature profile of the off-gases, the presence of a Hybrid Recuperator enhances the available work......-treated fuel then enters the anode side of the SOFC. Complete fuel oxidation is ensured in a burner by off-gases exiting the SOFC stacks. Off-gases are utilized as heat source for a SKC where a mixture of ammonia and water is expanded in a turbine to produce additional electric power. Thus, a triple novel......A hybrid plant that consists of a gasification system, Solid Oxide Fuel Cells (SOFC) and a Simple Kalina Cycle (SKC) is investigated. Woodchips are introduced into a fixed bed gasification plant to produce syngas, which is then fed into an integrated SOFC-SKC plant to produce electricity. The pre...

  13. Dual Pressure versus Hybrid Recuperation in an Integrated Solid Oxide Fuel Cell Cycle – Steam Cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2014-01-01

    A SOFC (solid oxide fuel cell) cycle running on natural gas was integrated with a ST (steam turbine) cycle. The fuel is desulfurized and pre-reformed before entering the SOFC. A burner was used to combust the remaining fuel after the SOFC stacks. The off-gases from the burner were used to produce...... pressure configuration steam cycle combined with SOFC cycle (SOFC-ST) was new and has not been studied previously. In each of the configuration, a hybrid recuperator was used to recovery the remaining energy of the off-gases after the HRSG. Thus, four different plants system setups were compared to each...... other to reveal the most superior concept with respect to plant efficiency and power. It was found that in order to increase the plant efficiency considerably, it was enough to use a single pressure with a hybrid recuperator instead of a dual pressure Rankine cycle....

  14. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.

    1993-01-01

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

  15. Integrated model of Korean spent fuel and high level waste disposal options - 16091

    International Nuclear Information System (INIS)

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21. century. The model addresses alternative design concepts for disposal of SNF of different types (Candu, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model's results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses. (authors)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. Reactivity management and burn-up management on JRR-3 silicide-fuel-core

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

    2007-08-01

    On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

  18. Evaluation of a Cogeneration Plant with Integrated Fuel Factory; Integrerad braenslefabrik med kraftvaermeanlaeggning - en utvaerdering

    Energy Technology Data Exchange (ETDEWEB)

    Atterhem, Lars

    2002-12-01

    A feasibility study was carried out in 1993 by Skellefteaa Kraft AB, to analyse the technical and economical possibilities to build a new baseload district heating production plant. The conclusion from the study was that, as a first step, a new cogeneration plant, based on a circulating fluidised bed boiler, should be built. The commissioning of the cogeneration plant took place in autumn 1996. The plant was prepared for a future integration with a biofuel drying process for pellets production. During spring 1996 an investment decision was taken and the fuel factory was erected in may 1997. Vaermeforsk Service AB has financed this research project and the Swedish state energy program (Fabel) has contributed with 33,7 Million SEK to the financing of the recovery electric power generation part of the fuel factory. The aim with this research project has been to evaluate and compare the integrated cogeneration plant fuel factory concept with a conventional co-generation plant, specially when it comes to increased power generation. The fuel factory comprises of fuel feeding system, fuel dryer, steam converter from fuel moisture to low pressure process steam, low pressure condensing turbine, cooling water system, fuel pellets production and storage with ship loading plant in the harbour of Skellefteaa. The steam to the fuel factory is extracted from the cogeneration turbine at a pressure level between 12-26 bar and the extraction flow has then already generated power in the cogeneration turbine. Power is also generated in the low pressure condensing turbine of the fuel factory. The low pressure steam is generated with fuel moisture in the steam converter. During the first years of operation there has been both conventional commissioning problems but also technical problems related to the new process concept. The last are for example corrosion and erosion problems, fouling problems of heat exchangers, capacity and leakage problems. The performance goals of the fuel

  19. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

    Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

  20. Dimensional measurements and eddy currents control of the sheath integrity for a set of irradiated candu fuel elements

    International Nuclear Information System (INIS)

    Gheorghe, G.; Man, I.

    2015-01-01

    During irradiation in the nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of sheath surface condition as well, which can lead to damages and even loss of integrity. This paper presents the results of dimensional measurements and of examination technique with eddy currents for three fuel elements of an irradiated CANDU fuel bundle. One of the fuel elements (FE), which is studied in detail, presented a crack about 40 mm long. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. This paper contains images of defects and interpretations of the causes of their occurrence. (authors)

  1. An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

    International Nuclear Information System (INIS)

    Swinhoe, Martyn T.; Tobin, Stephen J.; Fensin, Mike L.; Menlove, Howard O.

    2009-01-01

    be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC and A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full

  2. Further Improvement and System Integration of High Temperature Polymer Electrolyte Membrane Fuel Cells

    DEFF Research Database (Denmark)

    Li, Qingfeng; Jensen, Jens Oluf

    The new development in the field of polymer electrolyte membrane fuel cell (PEMFC) is high temperature PEMFC for operation above 100°C, which has been successfully demonstrated through the previous EC Joule III and the 5th framework programme. New challenges are encountered, bottlenecks for the new...... technology have been identified, and new concepts and solutions have been provisionally identified. FURIM is directed at tackling these key issues by concentrating on the further materials development, compatible technologies, and system integration of the high temperature PEMFC. The strategic developments...... of the FURIM are in three steps: (1) further improvement of the high temperature polymer membranes and related materials; (2) development of technological units including fuel cell stack, hydrocarbon reformer and afterburner, that are compatible with the HT-PEMFC; and (3) integration of the HT-PEMFC stack...

  3. INSIGHT: an integrated scoping analysis tool for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Noda, Hidefumi; Ito, Nobuaki; Maruyama, Taiji.

    1997-01-01

    An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis. (author)

  4. The need for integral critical experiments with low-moderated MOX fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The use of MOX fuel in commercial reactors is a means of burning plutonium originating from either surplus weapons or reprocessed irradiated uranium fuel. This requires the fabrication of MOX assemblies on an industrial scale. The OECD/NEA Expert Group on Experimental Needs for Criticality Safety has highlighted MOX fuel manufacturing, as an area in which there is a specific need for additional experimental data for validation purposes. Indeed, integral experiments with low-moderated MOX fuel are either scarce or not sufficiently accurate to provide an appropriate degree of validation of nuclear data and computer codes. New and accurate experimental data would enable a better optimisation of the fabrication process by decreasing the uncertainties in the determination of multiplication factors of configurations such as the homogenization of MOX powders. In this context, the OECD/NEA Nuclear Science Committee organised a workshop to address the following topics: expression and justification of the need for critical or near-critical experiments employing low-moderated MOX fuels; proposals for experimental programmes to address these needs; prospects for an international co-operative programme. The workshop was held at OECD headquarters in Paris on 14-15 April 2004. (author)

  5. Microcontroller based implementation of fuel cell and battery integrated hybrid power source

    International Nuclear Information System (INIS)

    Fahad, A.; Ali, S.M.; Bhatti, A.A.; Nasir, M

    2013-01-01

    This paper presents the implementation of a digitally controlled hybrid power source system, composed of fuel cell and battery. Use of individual fuel cell stacks as a power source, encounters many problems in achieving the desired load characteristics. A battery integrated, digitally controlled hybrid system is proposed for high pulse requirements. The proposed hybrid power source fulfils these peak demands with efficient flow of energy as compared to individual operations of fuel cell or battery system. A dc/dc converter is applied which provides an optimal control of power flow among fuel cell, battery and load. The proposed system efficiently overcomes the electrochemical constraints like over current, battery leakage current, and over and under voltage dips. By formulation of an intelligent algorithm and incorporating a digital technology (AVR Microcontroller), an efficient control is achieved over fuel cell current limit, battery charge, voltage and current. The hybrid power source is tested and analyzed by carrying out simulations using MATLAB simulink. Along with the attainment of desired complex load profiles, the proposed design can also be used for power enhancement and optimization for different capacities. (author)

  6. An integrated approach for investigation of failed nuclear fuel used at NPP Cernavoda Unit 1

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Popov, M.; Dobrin, R.; Staicu, C.

    1996-01-01

    At NPP Cernavoda-Unit 1 the fuel surveillance and the defect detection system in operation are based on monitoring the coolant activity concentration and on measuring the flux of delayed neutrons emitted by some short-lived fission products. In order to identify the failed fuel underwater non-destructive examination has to be performed. The major interest for the availability of underwater examination consists in the necessity of a speedy acquisition of the data on failed fuel in operation and of appropriate follow-up actions to be taken. Often the identification operation will be followed by more detailed examinations on selected fuel rods in the hot cells of the Post-irradiation Examination Laboratory of the Institute for Nuclear Research at Pitesti. Transfer of selected fuel rods will be done by the use of a type B(U) road transportation cask. Such an integrated approach will help to keep the level of activity concentration of the primary circuit well below the authorized limits. (author). 2 figs., 1 tab., 2 refs

  7. Development of Low Temperature Catalysts for an Integrated Ammonia PEM Fuel Cell

    OpenAIRE

    Hill, Alfred

    2014-01-01

    It is proposed that an integrated ammonia-PEM fuel cell could unlock the potential of ammonia to act as a high capacity chemical hydrogen storage vector and enable renewable energy to be delivered eectively to road transport applications. Catalysts are developed for low temperature ammonia decomposition with activity from 450 K (ruthenium and cesium on graphitised carbon nanotubes). Results strongly suggest that the cesium is present on the surface and close proximity to ruthenium nanoparticl...

  8. Integral logistics of the nuclear fuel Factory Juzbado; Logistica integral de la Fabrica de combustible Nulcear de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.

    2015-07-01

    The Logistic considers the complete process since the determination of possible demand, production planning, materials procurement, production control and delivery of final products to customer. This complete process is managed in all the scope under the same department called Planning and Logistic. This integration, some times really complex, has allowed to Enusa factory control all the key aspects that allow its running completely, considering the synergy's and important advantages to solve different problems. This article describes how we work of the main areas of procurement, production planning and control, fuel delivery and project planning of improvements on equipment's and factory systems, with an integrated management of all of them under the same direction. (Author)

  9. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  10. Integration of a molten carbonate fuel cell with a direct exhaust absorption chiller

    Science.gov (United States)

    Margalef, Pere; Samuelsen, Scott

    A high market value exists for an integrated high-temperature fuel cell-absorption chiller product throughout the world. While high-temperature, molten carbonate fuel cells are being commercially deployed with combined heat and power (CHP) and absorption chillers are being commercially deployed with heat engines, the energy efficiency and environmental attributes of an integrated high-temperature fuel cell-absorption chiller product are singularly attractive for the emerging distributed generation (DG) combined cooling, heating, and power (CCHP) market. This study addresses the potential of cooling production by recovering and porting the thermal energy from the exhaust gas of a high-temperature fuel cell (HTFC) to a thermally activated absorption chiller. To assess the practical opportunity of serving an early DG-CCHP market, a commercially available direct fired double-effect absorption chiller is selected that closely matches the exhaust flow and temperature of a commercially available HTFC. Both components are individually modeled, and the models are then coupled to evaluate the potential of a DG-CCHP system. Simulation results show that a commercial molten carbonate fuel cell generating 300 kW of electricity can be effectively coupled with a commercial 40 refrigeration ton (RT) absorption chiller. While the match between the two "off the shelf" units is close and the simulation results are encouraging, the match is not ideal. In particular, the fuel cell exhaust gas temperature is higher than the inlet temperature specified for the chiller and the exhaust flow rate is not sufficient to achieve the potential heat recovery within the chiller heat exchanger. To address these challenges, the study evaluates two strategies: (1) blending the fuel cell exhaust gas with ambient air, and (2) mixing the fuel cell exhaust gases with a fraction of the chiller exhaust gas. Both cases are shown to be viable and result in a temperature drop and flow rate increase of the

  11. Integration of a molten carbonate fuel cell with a direct exhaust absorption chiller

    Energy Technology Data Exchange (ETDEWEB)

    Margalef, Pere; Samuelsen, Scott [National Fuel Cell Research Center (NFCRC), University of California, Irvine, CA 92697-3550 (United States)

    2010-09-01

    A high market value exists for an integrated high-temperature fuel cell-absorption chiller product throughout the world. While high-temperature, molten carbonate fuel cells are being commercially deployed with combined heat and power (CHP) and absorption chillers are being commercially deployed with heat engines, the energy efficiency and environmental attributes of an integrated high-temperature fuel cell-absorption chiller product are singularly attractive for the emerging distributed generation (DG) combined cooling, heating, and power (CCHP) market. This study addresses the potential of cooling production by recovering and porting the thermal energy from the exhaust gas of a high-temperature fuel cell (HTFC) to a thermally activated absorption chiller. To assess the practical opportunity of serving an early DG-CCHP market, a commercially available direct fired double-effect absorption chiller is selected that closely matches the exhaust flow and temperature of a commercially available HTFC. Both components are individually modeled, and the models are then coupled to evaluate the potential of a DG-CCHP system. Simulation results show that a commercial molten carbonate fuel cell generating 300 kW of electricity can be effectively coupled with a commercial 40 refrigeration ton (RT) absorption chiller. While the match between the two ''off the shelf'' units is close and the simulation results are encouraging, the match is not ideal. In particular, the fuel cell exhaust gas temperature is higher than the inlet temperature specified for the chiller and the exhaust flow rate is not sufficient to achieve the potential heat recovery within the chiller heat exchanger. To address these challenges, the study evaluates two strategies: (1) blending the fuel cell exhaust gas with ambient air, and (2) mixing the fuel cell exhaust gases with a fraction of the chiller exhaust gas. Both cases are shown to be viable and result in a temperature drop and flow

  12. CPP-603 Underwater Fuel Storage Facility Site Integrated Stabilization Management Plan (SISMP), Volume I

    International Nuclear Information System (INIS)

    Denney, R.D.

    1995-10-01

    The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been constructed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1 to the Secretary of Energy regarding an improved schedule for remediation in the Defense Nuclear Facilities Complex. As described in the DNFSB Recommendation 94-1 Implementation Plan, issued February 28, 1995, an INEL Spent Nuclear Fuel Management Plan is currently under development to direct the placement of SNF currently in existing INEL facilities into interim storage, and to address the coordination of intrasite SNF movements with new receipts and intersite transfers that were identified in the DOE SNF Programmatic and INEL Environmental Restoration and Waste Management Environmental Impact Statement Record, of Decision. This SISMP will be a subset of the INEL Spent Nuclear Fuel Management Plan and the activities described are being coordinated with other INEL SNF management activities. The CPP-603 relocation activities have been assigned a high priority so that established milestones will be meet, but there will be some cases where other activities will take precedence in utilization of available resources. The Draft INEL Site Integrated Stabilization Management Plan (SISMP), INEL-94/0279, Draft Rev. 2, dated March 10, 1995, is being superseded by the INEL Spent Nuclear Fuel Management Plan and this CPP-603 specific SISMP

  13. Small PWR 'PFPWR50' using cermet fuel of Th-Pu particles

    International Nuclear Information System (INIS)

    Hirayama, Takashi; Shimazu, Yoichiro

    2009-01-01

    An innovative concept of PFPWR50 has been studied. The main feature of PFPWR50 has been to adopt TRISO coated fuel particles in a conventional PWR cladding. Coated fuel particle provides good confining ability of fission products. But it is pointed out that swelling of SiC layer at low temperature by irradiation has possibilities of degrading the integrity of coated fuel particle in the LWR environment. Thus, we examined the use of Cermet fuel replacing SiC layer to Zr metal or Zr compound. And the nuclear fuel has been used as fuel compact, which is configured to fix coated fuel particles in the matrix material to the shape of fuel pellet. In the previous study, graphite matrix is adopted as the matrix material. According to the burnup calculations of the several fuel concepts with those covering layers, we decide to use Zr layer embedded in Zr metal base or ZrC layer with graphite matrix. But carbon has the problem at low temperature by irradiation as well as SiC. Therefore, Zr covering layer and Zr metal base are finally selected. The other feature of PFPWR50 concept has been that the excess reactivity is suppressed during a cycle by initially loading burnable poison (gadolinia) in the fuels. In this study, a new loading pattern is determined by combining 7 types of assemblies in which the gadolinia concentration and the number of the fuel rods with gadolinia are different. This new core gives 6.7 equivalent full power years (EFPY) as the core life of a cycle. And the excess reactivity is suppressed to less than 2.0%Δk/k during the cycle. (author)

  14. Integrated hydrogen production process from cellulose by combining dark fermentation, microbial fuel cells, and a microbial electrolysis cell

    KAUST Repository

    Wang, Aijie; Sun, Dan; Cao, Guangli; Wang, Haoyu; Ren, Nanqi; Wu, Wei-Min; Logan, Bruce E.

    2011-01-01

    Hydrogen gas production from cellulose was investigated using an integrated hydrogen production process consisting of a dark fermentation reactor and microbial fuel cells (MFCs) as power sources for a microbial electrolysis cell (MEC). Two MFCs

  15. Welfare implications of the renewable fuel standard with an integrated tax-subsidy policy

    International Nuclear Information System (INIS)

    Skolrud, Tristan D.; Galinato, Gregmar I.

    2017-01-01

    This paper derives the optimal integrated tax-subsidy policy where one input is taxed and revenues are used to subsidize the use of a substitute input to reduce greenhouse gas emissions given the existing policies under the Renewable Fuel Standard policies. We measure the welfare effects and impact on cellulosic ethanol production after implementing the tax-subsidy policy using a general equilibrium model. A revenue-neutral integrated tax-subsidy scheme leads to a small positive tax rate for crude oil and a large positive subsidy for cellulosic ethanol because the former has a larger emissions coefficient than the latter. The overall welfare effects of an integrated tax subsidy scheme are less than a 1% increase for the economy but the growth in the cellulosic ethanol industry could range from 28% to 238% because the revenues from taxing crude oil are directly used to subsidize cellulosic ethanol production. - Highlights: • We derive an integrated tax-subsidy interacting with the Renewable Fuel Standard. • The policy is revenue-neutral. • Policy results in a small crude oil tax and a large cellulosic ethanol subsidy. • Simulations indicate a welfare-increasing optimal policy. • Growth in the cellulosic ethanol industry ranges from 28% to 238%.

  16. Reactivity determination of the Al2O3-B4C burnable poison as a function of its concentration in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Giada, Marino Reis

    2005-01-01

    Burnable poison rods made of Al 2 O 3 -B 4 C pellets with different concentrations of 10 B have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. The experiments evaluated the reactivity of the burnable poison rods as a function of the 10 B concentration, and the shadowing effect on the control rod reactivity worth as a function of the distance between the burnable position rods and the control rod. The results showed that the burnable poison rods have a non-linear behavior as function of the 10 B concentration, starting to reach an asymptotic value for concentrations higher than 7 g/cm 3 of 10 B. The shadowing effect on the control rods was substantial. When the burnable poison rods were beside the control rod, its reactivity worth decreased as much as 30 %, and when they were 10,5 cm distant, the control rod worth decreased by 7 %. The MCNP results for the burnable poison reactivity effects agreed within experimental errors with the measured values. (author)

  17. Thermochemical production of liquid fuels from biomass: Thermo-economic modeling, process design and process integration analysis

    International Nuclear Information System (INIS)

    Tock, Laurence; Gassner, Martin; Marechal, Francois

    2010-01-01

    A detailed thermo-economic model combining thermodynamics with economic analysis and considering different technological alternatives for the thermochemical production of liquid fuels from lignocellulosic biomass is presented. Energetic and economic models for the production of Fischer-Tropsch fuel (FT), methanol (MeOH) and dimethyl ether (DME) by means of biomass drying with steam or flue gas, directly or indirectly heated fluidized bed or entrained flow gasification, hot or cold gas cleaning, fuel synthesis and upgrading are reviewed and developed. The process is integrated and the optimal utility system is computed. The competitiveness of the different process options is compared systematically with regard to energetic, economic and environmental considerations. At several examples, it is highlighted that process integration is a key element that allows for considerably increasing the performance by optimal utility integration and energy conversion. The performance computations of some exemplary technology scenarios of integrated plants yield overall energy efficiencies of 59.8% (crude FT-fuel), 52.5% (MeOH) and 53.5% (DME), and production costs of 89, 128 and 113 Euro MWh -1 on fuel basis. The applied process design approach allows to evaluate the economic competitiveness compared to fossil fuels, to study the influence of the biomass and electricity price and to project for different plant capacities. Process integration reveals in particular potential energy savings and waste heat valorization. Based on this work, the most promising options for the polygeneration of fuel, power and heat will be determined in a future thermo-economic optimization.

  18. Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

    International Nuclear Information System (INIS)

    Garone, J.G.M.

    1983-01-01

    The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

  19. Control structure design of a solid oxide fuel cell and a molten carbonate fuel cell integrated system: Top-down analysis

    International Nuclear Information System (INIS)

    Jienkulsawad, Prathak; Skogestad, Sigurd; Arpornwichanop, Amornchai

    2017-01-01

    Highlights: • Control structure of the combined fuel cell system is designed. • The design target is trade-off between power generation and carbon dioxide emission. • Constraints are considered according to fuel cell safe operation. • Eight variables have to be controlled to maximize profit. • Two control structures are purposed for three active constraint regions. - Abstract: The integrated system of a solid oxide fuel cell and molten carbonate fuel cell theoretically has very good potential for power generation with carbon dioxide utilization. However, the control strategy of such a system needs to be considered for efficient operation. In this paper, a control structure design for an integrated fuel cell system is performed based on economic optimization to select manipulated variables, controlled variables and control configurations. The objective (cost) function includes a carbon tax to get an optimal trade-off between power generation and carbon dioxide emission, and constraints include safe operation. This study focuses on the top-down economic analysis which is the first part of the design procedure. Three actively constrained regions as a function of the main disturbances, namely, the fuel and steam feed rates, are identified; each region represents different sets of active constraints. Under nominal operating conditions, the system operates in region I. However, operating the fuel cell system in region I and II can use the same structure, but in region III, a different control structure is required.

  20. A Study on the Structural Integrity Issues of a Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Kang-Hee; Lee, Young-Ho; Yoon, Kyung-Ho; Kim, Jae-Yong; Song, Kun-Woo [Korea Atomic Energy Research Institute, 1045 Daedeokdaero Yuseong Daejeon 305-353 (Korea, Republic of)

    2009-06-15

    A dual-cooled fuel rod has an internal coolant flow passage in addition to the external one. A remarkable power up-rate can be achieved due to the increased surface area, which may draw great interests from the fuel researchers, designers and vendors. However, it requires effective resolution to the difficult technical issues when a fuel assembly is to be realized. It becomes much more difficult if a tough boundary condition needs to be satisfied such as a compatibility with the existing reactor internal structures. This kind of challenge is tackled through a national R and D project in Korea: to develop the structural components of a dual-cooled fuel that should be compatible with the current OPR 1000 (Korea Standard Nuclear Power Plant) internal structures. Fuel rod supporting structures, top and bottom end pieces and guide tubes are the components. Besides, the fuel rod components have to be developed as well since the fuel rod's geometry becomes much different from the conventional rod's one. The dimension change may well affect the above mentioned structural components. As a part of the work, structural integrity of the components of a dual-cooled fuel rod is studied in this paper. The investigated topics are: i) the thickness determination of a cladding tube (especially outer tube of a large diameter), ii) vibration issue of an inner cladding tube, iii) design concern of plenum spring and spacer. The cladding thickness issue arises due to the increased outside diameter of a fuel rod, which is caused by an internal flow passage formation. Among the criteria for the thickness determination, an elastic buckling criteria was focused on. Theoretical background for the well-known formula (such as a stability problem) was revisited. Verification tests were carried out independently with using a cladding tube of PHWR fuel rod. Results showed that the formula was not conservative to apply for the cladding thickness determination. Minimum thickness for the

  1. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  2. Engine-integrated solid oxide fuel cells for efficient electrical power generation on aircraft

    Science.gov (United States)

    Waters, Daniel F.; Cadou, Christopher P.

    2015-06-01

    This work investigates the use of engine-integrated catalytic partial oxidation (CPOx) reactors and solid oxide fuel cells (SOFCs) to reduce fuel burn in vehicles with large electrical loads like sensor-laden unmanned air vehicles. Thermodynamic models of SOFCs, CPOx reactors, and three gas turbine (GT) engine types (turbojet, combined exhaust turbofan, separate exhaust turbofan) are developed and checked against relevant data and source material. Fuel efficiency is increased by 4% and 8% in the 50 kW and 90 kW separate exhaust turbofan systems respectively at only modest cost in specific power (8% and 13% reductions respectively). Similar results are achieved in other engine types. An additional benefit of hybridization is the ability to provide more electric power (factors of 3 or more in some cases) than generator-based systems before encountering turbine inlet temperature limits. A sensitivity analysis shows that the most important parameters affecting the system's performance are operating voltage, percent fuel oxidation, and SOFC assembly air flows. Taken together, this study shows that it is possible to create a GT-SOFC hybrid where the GT mitigates balance of plant losses and the SOFC raises overall system efficiency. The result is a synergistic system with better overall performance than stand-alone components.

  3. Integrated Data Base for 1989: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1989-11-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1988. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, commercial reactor and fuel cycle facility decommissioning waste, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous, highly radioactive materials that may require geologic disposal. 45 figs., 119 tabs

  4. Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  6. Advanced control system for the Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

    1993-01-01

    A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface

  7. A software tool integrated risk assessment of spent fuel transpotation and storage

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Mi Rae; Almomani, Belal; Ham, Jae Hyun; Kang, Hyun Gook [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Christian, Robby [Dept. of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy (Korea, Republic of); Kim, Bo Gyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of)

    2017-06-15

    When temporary spent fuel storage pools at nuclear power plants reach their capacity limit, the spent fuel must be moved to an alternative storage facility. However, radioactive materials must be handled and stored carefully to avoid severe consequences to the environment. In this study, the risks of three potential accident scenarios (i.e., maritime transportation, an aircraft crashing into an interim storage facility, and on-site transportation) associated with the spent fuel transportation process were analyzed using a probabilistic approach. For each scenario, the probabilities and the consequences were calculated separately to assess the risks: the probabilities were calculated using existing data and statistical models, and the consequences were calculated using computation models. Risk assessment software was developed to conveniently integrate the three scenarios. The risks were analyzed using the developed software according to the shipment route, building characteristics, and spent fuel handling environment. As a result of the risk analysis with varying accident conditions, transportation and storage strategies with relatively low risk were developed for regulators and licensees. The focus of this study was the risk assessment methodology; however, the applied model and input data have some uncertainties. Further research to reduce these uncertainties will improve the accuracy of this mode.

  8. Integrated data base for 1990: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1990-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1989. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 22 refs., 48 figs., 109 tabs

  9. Integrated Data Base for 1991: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1991-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1990. These data are based on the most reliable information available form government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated generally through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 160 refs., 61 figs., 142 tabs

  10. A software tool integrated risk assessment of spent fuel transpotation and storage

    International Nuclear Information System (INIS)

    Yun, Mi Rae; Almomani, Belal; Ham, Jae Hyun; Kang, Hyun Gook; Christian, Robby; Kim, Bo Gyung; Lee, Sang Hoon

    2017-01-01

    When temporary spent fuel storage pools at nuclear power plants reach their capacity limit, the spent fuel must be moved to an alternative storage facility. However, radioactive materials must be handled and stored carefully to avoid severe consequences to the environment. In this study, the risks of three potential accident scenarios (i.e., maritime transportation, an aircraft crashing into an interim storage facility, and on-site transportation) associated with the spent fuel transportation process were analyzed using a probabilistic approach. For each scenario, the probabilities and the consequences were calculated separately to assess the risks: the probabilities were calculated using existing data and statistical models, and the consequences were calculated using computation models. Risk assessment software was developed to conveniently integrate the three scenarios. The risks were analyzed using the developed software according to the shipment route, building characteristics, and spent fuel handling environment. As a result of the risk analysis with varying accident conditions, transportation and storage strategies with relatively low risk were developed for regulators and licensees. The focus of this study was the risk assessment methodology; however, the applied model and input data have some uncertainties. Further research to reduce these uncertainties will improve the accuracy of this mode

  11. Integrated anode structure for passive direct methanol fuel cells with neat methanol operation

    Science.gov (United States)

    Wu, Huijuan; Zhang, Haifeng; Chen, Peng; Guo, Jing; Yuan, Ting; Zheng, Junwei; Yang, Hui

    2014-02-01

    A microporous titanium plate based integrated anode structure (Ti-IAS) suitable for passive direct methanol fuel cells (DMFCs) fueled with neat methanol is reported. This anode structure incorporates a porous titanium plate as a methanol mass transfer barrier and current collector, pervaporation film for passively vaporizing methanol, vaporous methanol cavity for evenly distributing fuel, and channels for carbon dioxide venting. With the effective control of methanol delivery rate, the Ti-IAS based DMFC allows the direct use of neat methanol as the fuel source. In the meantime, the required water for methanol-oxidation reaction at the anode can also be fully recovered from the cathode with the help of the highly hydrophobic microporous layer in the cathode. DMFCs incorporating this new anode structure exhibit a power density as high as 40 mW cm-2 and a high volumetric energy density of 489 Wh L-1 operating with neat methanol and at 25 °C. Importantly, no obvious performance degradation of the passive DMFC system is observed after more than 90 h of continuous operation. The experimental results reveal that the compact DMFC based on the Ti-IAS exhibits a substantial potential as power sources for portable applications.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  13. Design considerations for a 10-KW integrated hydrogen-oxygen regenerative fuel cell system

    International Nuclear Information System (INIS)

    Hoberecht, M.A.; Gonzalez-Sanabria, O.D.; Miller, T.B.; Rieker, L.L.

    1984-01-01

    Integration of an alkaline fuel cell subsystem with an alkaline electrolysis subsystem to form a regenerative fuel cell (RFC) system for low-earth-orbit (LEO) applications characterized by relatively high overall round-trip electrical efficiency, long life, and high reliability is possible with present state-of-the-art technology. A hypothetical 10-kW system is being computer modeled and studied based on data from ongoing contractual efforts in both the alkaline fuel cell and alkaline water electrolysis areas. The alkaline fuel cell technology is being developed under an NASA-LeRC program with United Technologies Corporation (UTC), utilizing advanced cell components and standard Shuttle-Orbiter system hardware. The alkaline electrolysis technology is that of Life Systems, Inc. (LSI), which uses a static water vapor feed technique and scaled-up cell hardware being developed under an NASA-LeRC program. This paper addresses the computeraided study of the performance, operating, and design parameters of the hypothetical system

  14. Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S-S; Kim, S-H; Jung, Y-K; Yang, C-Y; Kim, I-G; Choi, Y-H; Kim, H-J; Kim, M-W; Rho, B-H [KINS, Daejeon (Korea, Republic of)

    2008-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is the final report.

  15. Study on the standard establishment for the integrity assessment of nuclear fuel cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. S.; Kim, S. H.; Jung, Y. K.; Yang, C. Y.; Kim, I. G.; Choi, Y. H.; Kim, H. J.; Kim, M. W.; Rho, B. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is 2nd term report.

  16. Integrated monitoring and reviewing systems for the Rokkasho Spent Fuel Receipt and Storage Facility

    International Nuclear Information System (INIS)

    Yokota, Yasuhiro; Ishikawa, Masayuki; Matsuda, Yuji

    1998-01-01

    The Rokkasho Spent Fuel Receipt and Storage (RSFS) Facility at the Rokkasho Reprocessing Plant (RRP) in Japan is expected to begin operations in 1998. Effective safeguarding by International Atomic Energy Agency (IAEA) and Japan Atomic Energy Bureau (JAEB) inspectors requires monitoring the time of transfer, direction of movement, and number of spent fuel assemblies transferred. At peak throughput, up to 1,000 spent fuel assemblies will be accepted by the facility in a 90-day period. In order for the safeguards inspector to efficiently review the resulting large amounts of inspection information, an unattended monitoring system was developed that integrates containment and surveillance (C/S) video with radiation monitors. This allows for an integrated review of the facility's radiation data, C/S video, and operator declaration data. This paper presents an outline of the integrated unattended monitoring hardware and associated data reviewing software. The hardware consists of a multicamera optical surveillance (MOS) system radiation monitoring gamma-ray and neutron detector (GRAND) electronics, and an intelligent local operating network (ILON). The ILON was used for time synchronization and MOS video triggers. The new software consists of a suite of tools, each one specific to a single data type: radiation data, surveillance video, and operator declarations. Each tool can be used in a stand-alone mode as a separate ion application or configured to communicate and match time-synchronized data with any of the other tools. A data summary and comparison application (Integrated Review System [IRS]) coordinates the use of all of the data-specific review tools under a single-user interface. It therefore automates and simplifies the importation of data and the data-specific analyses

  17. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Nunez C, A.

    2003-01-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  18. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  19. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  20. Study of renewable energy, fuel cell and demotics integration for stationary energy production

    Energy Technology Data Exchange (ETDEWEB)

    Andaloro, L.; Ferraro, M.; Sergi, F.; Brunaccini, G.; Antonucci, V. [National Research Inst., Messina (Italy)

    2009-07-01

    This paper described a study in which a small house equipped with various renewable technologies was modelled. The aim of the study was to evaluated the integration of fuel cells with various other energy sources. Technologies installed in the house included a photovoltaic (PV) system; a hydrogen system; fuel cells; a battery-storage system; and a thermal solar panel. Maximum energy savings were evaluated for different configurations and combinations of the installed energy sources. A domotic system was also used to automatically control the use of electrical appliances and improve safety and comfort. An energy side management system was designed and compared with a demand side management system. Various scenarios were simulated in order to test the energy management systems in relation to the automated domotic system.

  1. A description of the demonstration Integral Fast Reactor fuel cycle facility

    International Nuclear Information System (INIS)

    Courtney, J.C.; Carnes, M.D.; Dwight, C.C.; Forrester, R.J.

    1991-01-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations

  2. Fuel cycle facility control system for the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Benedict, R.W.; Tate, D.A.

    1993-01-01

    As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations

  3. Data processing in the integrated data base for spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Morrison, G.W.; Notz, K.J.

    1984-01-01

    The Integrated Data Base (IDB) Program at Oak Ridge National Laboratory (ORNL) produces for the U.S. Department of Energy (DOE) the official spent fuel and radioactive waste inventories and projections for the United States through the year 2020. Inventory data are collected and checked for consistency, projection data are calculated based on specified assumptions, and both are converted to a standard format. Spent fuel and waste radionclides are decayed as a function of time. The resulting information constitutes the core data files called the Past/Present/Future (P/P/F) data base. A data file management system, SAS /sup R/, is used to retrieve the data and create several types of output: an annual report, an electronic summary data file designed for IBM-PC /sup R/ -compatible computers, and special-request reports

  4. Micro direct methanol fuel cell with perforated silicon-plate integrated ionomer membrane

    DEFF Research Database (Denmark)

    Larsen, Jackie Vincent; Dalslet, Bjarke Thomas; Johansson, Anne-Charlotte Elisabeth Birgitta

    2014-01-01

    This article describes the fabrication and characterization of a silicon based micro direct methanol fuel cell using a Nafion ionomer membrane integrated into a perforated silicon plate. The focus of this work is to provide a platform for micro- and nanostructuring of a combined current collector...... at a perforation ratio of 40.3%. The presented fuel cells also show a high volumetric peak power density of 2 mW cm−3 in light of the small system volume of 480 μL, while being fully self contained and passively feed....... and catalytic electrode. AC impedance spectroscopy is utilized alongside IV characterization to determine the influence of the plate perforation geometries on the cell performance. It is found that higher ratios of perforation increases peak power density, with the highest achieved being 2.5 mW cm−2...

  5. HT-PEM Fuel Cell System with Integrated Thermoelectric Exhaust Heat Recovery

    DEFF Research Database (Denmark)

    Gao, Xin

    This thesis presents two case studies on improving the efficiency and the loadfollowing capability of a high temperature polymer electrolyte membrane (HTPEM) fuel cell system by the application of thermoelectric (TE) devices. TE generators (TEGs) are harnessed to recover the system exhaust gas...... developed three-dimensional numerical model in ANSYS Fluent®. This thesis introduces the progress of this project in a cognitive order. The first chapter initially prepares the theory and characteristics of the fuel cell system and TE devices. Project motivations are conceived. Then similar studies existing...... power output on the subsystem design and performance were also systematically analyzed. The TEG subsystem configuration is optimized. The usefulness and convenience of the model are proved. TE coolers (TECs) are integrated into the methanol evaporator of the HT-PEM system for improving the whole system...

  6. Spent fuel and radioactive waste: an integrated data base of inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Notz, K.J.; Forsberg, C.W.; Mastal, E.F.

    1984-01-01

    The Integrated Data Base (IDB) Program provides official US Department of Energy (DOE) data on spent fuel and radioactive waste inventories, projections, and characteristics. This information is provided through the cooperative efforts of the IDB Program and DOE lead offices, lead sites, major programs, and generator sites. The program is entering its fifth year, and major accomplishments are summarized in three broad areas: (1) the annual inventory report, including ORIGEN2 applications and a Quality Assurance (QA) plan; (2) the summary data file and direct user access; and (3) data processing methodology and support to other programs. Plans for future work in these areas are outlined briefly, including increased utilization of personal computers. Some examples of spent fuel data are given in terms of projected quantities for two growth scenarios, burnup and age profile of the existing inventory, and the approximate specific thermal power relative to high-level waste (HLW) from various sources. 4 refs., 2 figs., 3 tabs

  7. Fuels planning: science synthesis and integration; social issues fact sheet 13: Strategies for managing fuels and visual quality

    Science.gov (United States)

    Christine Esposito

    2006-01-01

    The public's acceptance of forest management practices, including fuels reduction, is heavily based on how forests look. Fuels managers can improve their chances of success by considering aesthetics when making management decisions. This fact sheet reviews a three-part general strategy for managing fuels and visual quality: planning, implementation, and monitoring...

  8. Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 05: fuel treatment principles for complex landscapes

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    Appropriate types of thinning and surface fuel treatments are clearly useful in reducing surface and crown fire hazards under a wide range of fuels and topographic situations. This paper provides well-established scientific principles and simulation tools that can be used to adjust fuel treatments to attain specific risk levels.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  10. 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)

  11. Storage, handling and movement of fuel and related components at nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

  12. Biomass gasification integrated with a solid oxide fuel cell and Stirling engine

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2014-01-01

    An integrated gasification solid oxide fuel cell (SOFC) and Stirling engine for combined heat and power application is analyzed. The target for electricity production is 120 kW. Woodchips are used as gasification feedstock to produce syngas, which is then used to feed the SOFC stacks...... for electricity production. Unreacted hydrocarbons remaining after the SOFC are burned in a catalytic burner, and the hot off-gases from the burner are recovered in a Stirling engine for electricity and heat production. Domestic hot water is used as a heat sink for the Stirling engine. A complete balance...

  13. Integration of Fuel Cell Micro-CHPs on Low. Voltage Grid: A Danish Case Study

    DEFF Research Database (Denmark)

    You, Shi; Marra, Francesco; Træholt, Chresten

    2012-01-01

    The future significance of fuel cell (FC) powered micro combined heat and power (micro-CHP) units in meeting the residential energy demands is set to increase, which may have a considerable impact on the low voltage (LV) grid. The objective of this paper is to investigate into the related technical...... issues using a Danish case study with different penetration levels of uncoordinated FC micro-CHPs. Based on the findings, it is recommended to design grid oriented integration strategies such as Virtual Power Plants (VPPs) for achieving future smart grids with a large roll out of distributed energy...

  14. Effect of humic acids on electricity generation integrated with xylose degradation in microbial fuel cells

    DEFF Research Database (Denmark)

    Huang, Liping; Angelidaki, Irini

    2008-01-01

    Pentose and humic acids (HA) are the main components of hydrolysates, the liquid fraction produced during thermohydrolysis of lignocellulosic material. Electricity generation integrated with xylose (typical pentose) degradation as well as the effect of HA on electricity production in microbial fuel...... to controls where HAs were not added, addition of commercial HA resulted in increase of power density and coulombic efficiency, which ranged from 7.5% to 67.4% and 24% to 92.6%, respectively. Digested manure wastewater (DMW) was tested as potential mediator for power generation due to its content of natural...

  15. Implementation of a Gadolinium Burnable Absorber in the Carbide LEU-NTR

    International Nuclear Information System (INIS)

    Venneria, Paolo; Kim, Yonghee

    2015-01-01

    Among the most crucial are the rapid reactivity depletion during full-power operation and the positive reactivity insertion during the full-submersion criticality accident. In previous work, it has been suggested that both challenges can be mitigated through the successful implementation of a burnable absorber in the active core. Of the poisons previously surveyed, one of the most promising is Gadolinium in the form of Gadolina (Gd2O4). This paper explores the possibility of different methods by which the Gadolinia can be implemented in the core and makes a preliminary study of its effect on the full submersion criticality accident and the reactivity depletion during operation. The application of a Gadolinium neutron absorber in the active core region of the LEU-NTR has been shown to be neutronically feasible. It can be introduced into the core in various locations without resulting in core performance loss. The utility of the poison in terms of mitigating the full-submersion reactivity accident and the rapid change in reactivity during full-power operation have been preliminarily shown and the first steps towards eventual implementation made. Future work will consist of determining the maximum poison content in the core and tailoring the self-shielding effect in order to determine a specific Gd depletion rate

  16. Dynamic modeling, experimental evaluation, optimal design and control of integrated fuel cell system and hybrid energy systems for building demands

    Science.gov (United States)

    Nguyen, Gia Luong Huu

    Fuel cells can produce electricity with high efficiency, low pollutants, and low noise. With the advent of fuel cell technologies, fuel cell systems have since been demonstrated as reliable power generators with power outputs from a few watts to a few megawatts. With proper equipment, fuel cell systems can produce heating and cooling, thus increased its overall efficiency. To increase the acceptance from electrical utilities and building owners, fuel cell systems must operate more dynamically and integrate well with renewable energy resources. This research studies the dynamic performance of fuel cells and the integration of fuel cells with other equipment in three levels: (i) the fuel cell stack operating on hydrogen and reformate gases, (ii) the fuel cell system consisting of a fuel reformer, a fuel cell stack, and a heat recovery unit, and (iii) the hybrid energy system consisting of photovoltaic panels, fuel cell system, and energy storage. In the first part, this research studied the steady-state and dynamic performance of a high temperature PEM fuel cell stack. Collaborators at Aalborg University (Aalborg, Denmark) conducted experiments on a high temperature PEM fuel cell short stack at steady-state and transients. Along with the experimental activities, this research developed a first-principles dynamic model of a fuel cell stack. The dynamic model developed in this research was compared to the experimental results when operating on different reformate concentrations. Finally, the dynamic performance of the fuel cell stack for a rapid increase and rapid decrease in power was evaluated. The dynamic model well predicted the performance of the well-performing cells in the experimental fuel cell stack. The second part of the research studied the dynamic response of a high temperature PEM fuel cell system consisting of a fuel reformer, a fuel cell stack, and a heat recovery unit with high thermal integration. After verifying the model performance with the

  17. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  18. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  19. Integrated data base for 1988: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1988-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1987. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reportd for miscellaneous, highly radioactive materials that may require geologic disposal. 89 refs., 46 figs., 104 tabs

  20. Integrated data base for 1986: spent fuel and radioactive waste inventories, projections, and characteristics. Revision 2

    International Nuclear Information System (INIS)

    1986-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US Department of Energy (DOE) radioactive wastes through December 31, 1985, based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. Current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the expected defense-related and private industrial and institutional activities and the latest DOE/Energy Information Administration (EIA) projections of US commercial nuclear power growth. The materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or calculated isotopic compositions

  1. Energy consumption analysis of integrated flowsheets for production of fuel ethanol from lignocellulosic biomass

    International Nuclear Information System (INIS)

    Cardona Alzate, C.A.; Sanchez Toro, O.J.

    2006-01-01

    Fuel ethanol is considered one of the most important renewable fuels due to the economic and environmental benefits of its use. Lignocellulosic biomass is the most promising feedstock for producing bioethanol due to its global availability and to the energy gain that can be obtained when non-fermentable materials from biomass are used for cogeneration of heat and power. In this work, several process configurations for fuel ethanol production from lignocellulosic biomass were studied through process simulation using Aspen Plus. Some flowsheets considering the possibilities of reaction-reaction integration were taken into account among the studied process routes. The flowsheet variants were analyzed from the energy point of view utilizing as comparison criterion the energy consumption needed to produce 1 L of anhydrous ethanol. Simultaneous saccharification and cofermentation process with water recycling showed the best results accounting an energy consumption of 41.96 MJ/L EtOH. If pervaporation is used as dehydration method instead of azeotropic distillation, further energy savings can be obtained. In addition, energy balance was estimated using the results from the simulation and literature data. A net energy value of 17.65-18.93 MJ/L EtOH was calculated indicating the energy efficiency of the lignocellulosic ethanol

  2. Integrated production of merchantable wood and wood fuels in industry; Teollisuuden ainespuun ja puupolttoaineen integroitu tuotanto

    Energy Technology Data Exchange (ETDEWEB)

    Kuvaja, K [Enso Oy, Imatra (Finland). Forest Dept.

    1997-12-01

    The aim of this project is the economically profitable integrated harvesting of industrial wood and firewood especially in harvesting of small-diameter first thinning wood. The research in 1994 was concentrated on improvement of the quality of the chipping methods based on chain-flail debarking chipping method, and on determination of the possible utilisation targets for the fuel fraction. A reasonably large drum debarking test was also carried out at the industrial scale debarking station of the Enocell Oy. More than 80 000 m{sup 3} of first thinning wood was delivered by Enocell during this project. The quality of wood chips, produced using the chain-flail delimbing method, could be improved in the case of pine nearly to the required quality level, but additional measures are still needed in the case of birch. The fuel fraction deliveries to different points of utilisation was started. The particle size of the fuel fraction appeared to be good after crushing. In 1995 a chain-flail-drum debarking chipping unit was developed to improve and homogenise the quality of chips. (orig.)

  3. Improved method to demonstrate the structural integrity of high density fuel storage racks

    International Nuclear Information System (INIS)

    Hinderks, M.; Ungoreit, H.; Kremer, G.

    2001-01-01

    Reracking of existing fuel pools to the maximum extent is desirable from an economical point of view. This goal can be achieved by minimizing the gaps between the spent fuel storage racks. Since the rack design is aimed at enabling consolidated fuel rod storage, additional requirements arise with respect to the design and the structural analysis. The loads resulting from seismic events are decisive for the structural analysis and require a specially detailed and in-depth analysis for high seismic loads. The verification of structural integrity and functionality is performed in two phases. In the first phase the motional behavior of single racks, rows of racks and, where required, of all racks in the pool is simulated by excitation with displacement time histories under consideration of the fluid-structure interaction (FSI). The displacements from these simulations are evaluated, while the loads are utilized as input data for the structural analysis of the racks and the pool floor. The structural analyses for the racks comprise substantially stress analyses for base material and welds as well as stability analyses for the support channels and the rack outside walls. The analyses are performed in accordance with the specified codes and standards

  4. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

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

    International Nuclear Information System (INIS)

    Orechwa, Y.; Bucher, R.G.

    1994-01-01

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

  6. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

  7. Energy consumption analysis of integrated flowsheets for production of fuel ethanol from lignocellulosic biomass

    Energy Technology Data Exchange (ETDEWEB)

    Cardona Alzate, C.A. [Department of Chemical Engineering, National University of Colombia at Manizales, Cra. 27 No. 64-60, Manizales (Colombia)]. E-mail: ccardonaal@unal.edu.co; Sanchez Toro, O.J. [Department of Chemical Engineering, National University of Colombia at Manizales, Cra. 27 No. 64-60, Manizales (Colombia); Department of Engineering, University of Caldas, Calle 65 No. 26-10, Manizales (Colombia)

    2006-10-15

    Fuel ethanol is considered one of the most important renewable fuels due to the economic and environmental benefits of its use. Lignocellulosic biomass is the most promising feedstock for producing bioethanol due to its global availability and to the energy gain that can be obtained when non-fermentable materials from biomass are used for cogeneration of heat and power. In this work, several process configurations for fuel ethanol production from lignocellulosic biomass were studied through process simulation using Aspen Plus. Some flowsheets considering the possibilities of reaction-reaction integration were taken into account among the studied process routes. The flowsheet variants were analyzed from the energy point of view utilizing as comparison criterion the energy consumption needed to produce 1 L of anhydrous ethanol. Simultaneous saccharification and cofermentation process with water recycling showed the best results accounting an energy consumption of 41.96 MJ/L EtOH. If pervaporation is used as dehydration method instead of azeotropic distillation, further energy savings can be obtained. In addition, energy balance was estimated using the results from the simulation and literature data. A net energy value of 17.65-18.93 MJ/L EtOH was calculated indicating the energy efficiency of the lignocellulosic ethanol.

  8. Fuel cell-powered microfluidic platform for lab-on-a-chip applications: Integration into an autonomous amperometric sensing device.

    Science.gov (United States)

    Esquivel, J P; Colomer-Farrarons, J; Castellarnau, M; Salleras, M; del Campo, F J; Samitier, J; Miribel-Català, P; Sabaté, N

    2012-11-07

    The present paper reports for the first time the integration of a microfluidic system, electronics modules, amperometric sensor and display, all powered by a single micro direct methanol fuel cell. In addition to activating the electronic circuitry, the integrated power source also acts as a tuneable micropump. The electronics fulfil several functions. First, they regulate the micro fuel cell output power, which off-gas controls the flow rate of different solutions toward an electrochemical sensor through microfluidic channels. Secondly, as the fuel cell powers a three-electrode electrochemical cell, the electronics compare the working electrode output signal with a set reference value. Thirdly, if the concentration measured by the sensor exceeds this threshold value, the electronics switch on an integrated organic display. This integrated approach pushes forward the development of truly autonomous point-of-care devices relying on electrochemical detection.

  9. Dynamic modelling and characterisation of a solid oxide fuel cell integrated in a gas turbine cycle

    Energy Technology Data Exchange (ETDEWEB)

    Thorud, Bjoern

    2005-07-01

    This thesis focuses on three main areas within the field of SOFC/GT-technology: 1) Development of a dynamic SOFC/GT model. 2) Model calibration and sensitivity study. 3) Assessment of the dynamic properties of a SOFC/GT power plant. The SOFC/GT model developed in this thesis describes a pressurised tubular Siemens Westinghouse-type SOFC, which is integrated in a gas turbine cycle. The process further includes a plate-fin recuperator for stack air preheating, a prereformer, an anode exhaust gas recycling loop for steam/carbon-ratio control, an afterburner and a shell-tube heat exchanger for air preheating. The fuel cell tube, the recuperator and the shell-tube heat exchanger are spatially distributed models. The SOFC model is further thermally integrated with the prereformer. The compressor and turbine models are based on performance maps as a general representation of the characteristics. In addition, a shaft model which incorporates moment of inertia is included to account for gas turbine transients. The SOFC model is calibrated against experimentally obtained data from a single-cell experiment performed on a Siemens Westinghouse tubular SOFC. The agreement between the model and the experimental results is good. The sensitivity study revealed that the degree of prereforming is of great importance with respect to the axial temperature distribution of the fuel cell. Types of malfunctions are discussed prior to the dynamic behaviour study. The dynamic study of the SOFC/GT process is performed by simulating small and large load changes according to three different strategies; 1) Load change at constant mean fuel cell temperature. 2) Load change at constant turbine inlet temperature. 3) Load change at constant shaft speed. Of these three strategies, the constant mean fuel cell temperature strategy appears to be the most rapid load change method. Furthermore, this strategy implies the lowest degree of thermal cycling, the smoothest fuel cell temperature distribution and

  10. Process integration and optimization of a solid oxide fuel cell – Gas turbine hybrid cycle fueled with hydrothermally gasified waste biomass

    International Nuclear Information System (INIS)

    Facchinetti, Emanuele; Gassner, Martin; D’Amelio, Matilde; Marechal, François; Favrat, Daniel

    2012-01-01

    Due to its suitability for using wet biomass, hydrothermal gasification is a promising process for the valorization of otherwise unused waste biomass to synthesis gas and biofuels. Solid oxide fuel cell (SOFC) based hybrid cycles are considered as the best candidate for a more efficient and clean conversion of (bio) fuels. A significant potential for the integration of the two technologies is expected since hydrothermal gasification requires heat at 673–773 K, whereas SOFC is characterized by heat excess at high temperature due to the limited electrochemical fuel conversion. This work presents a systematic process integration and optimization of a SOFC-gas turbine (GT) hybrid cycle fueled with hydrothermally gasified waste biomass. Several design options are systematically developed and compared through a thermodynamic optimization approach based on First Law and exergy analysis. The work demonstrates the considerable potential of the system that allows for converting wet waste biomass into electricity at a First Law efficiency of up to 63%, while simultaneously enabling the separation of biogenic carbon dioxide for further use or sequestration. -- Highlights: ► Hydrothermal gasification is a promising process for the valorization of waste wet biomass. ► Solid Oxide Fuel Cell – Gas Turbine hybrid cycle emerges as the best candidates for conversion of biofuels. ► A systematic process integration and optimization of a SOFC-GT hybrid cycle fuelled with hydrothermally gasified biomass is presented. ► The system may convert wet waste biomass to electricity at a First Law efficiency of 63% while separating the biogenic carbon dioxide. ► The process integration enables to improve the First Law efficiency of around 4% with respect to a non-integrated system.

  11. Further Improvement and System Integration of High Temperature Polymer Electrolyte Membrane Fuel Cells

    DEFF Research Database (Denmark)

    Jensen, Jens Oluf; Li, Qingfeng

    Polymer electrolyte membrane fuel cell (PEMFC) technology based on Nafion membranes can operate at temperatures around 80°C. The new development in the field is high temperature PEMFC for operation above 100°C, which has been successfully demonstrated through the previous EC Joule III and the 5th......, and system integration of the high temperature PEMFC. The strategic developments of the FURIM are in three steps: (1) further improvement of the high temperature polymer membranes and related materials; (2) development of technological units including fuel cell stack, hydrocarbon reformer, afterburner...... and power management system, that are compatible with the HT-PEMFC; and (3) integration of the HT-PEMFC stack with these compatible subunits. The main goal of the project is a 2kWel HT-PEMFC stack operating in a temperature range of 120-220°C, with a single cell performance target of 0.7 A/cm² at a cell...

  12. Application of the integral method to modelling the oxidation of defected fuel elements

    International Nuclear Information System (INIS)

    Kolar, M.

    1995-06-01

    The starting point for this report is the discrepancy reported in previous work between the reaction-diffusion calculations and the CEX-1 experiment, which involves storage of defected fuel elements in air at 150 deg C. This discrepancy is considerably diminished here by a more critical choice of theoretical parameters, and by taking into account the fact that different CEX-1 fuel elements were oxidized at very different rates and that the fuel element used previously for comparison with theoretical calculations actually underwent two limited-oxygen-supply cycles. Much better agreement is obtained here between the theory and the third, unlimited-air, storage period of the CEX-1 experiment. The approximate integral method is used extensively for the solution of the one-dimensional diffusion moving-boundary problems that may describe various storage periods of the CEX-1 experiment. In some cases it is easy to extend this method to arbitrary precision by using higher moments of the diffusion equation. Using this method, the validity of quasi-steady-state approximation is verified. Diffusion-controlled oxidation is also studied. In this case, for the unlimited oxygen supply, the integral method leads to an exact analytical solution for linear geometry, and to a good analytical approximation of the solution for the spherically symmetric geometry. These solutions may have some application in the analysis of experiments on the oxidation of small UO 2 fragments or powders when the individual UO 2 grains may be considered to be approximately spherical. (author). 23 refs., 5 tabs., 11 figs

  13. Thermodynamic performance analysis of a fuel cell trigeneration system integrated with solar-assisted methanol reforming

    International Nuclear Information System (INIS)

    Wang, Jiangjiang; Wu, Jing; Xu, Zilong; Li, Meng

    2017-01-01

    Highlights: • Propose a fuel cell trigeneration system integrated with solar-assisted methanol reforming. • Optimize the reaction parameters of methanol steam reforming. • Present the energy and exergy analysis under design and off-design work conditions. • Analyze the contributions of solar energy to the trigeneration system. - Abstract: A solar-assisted trigeneration system for producing electricity, cooling, and heating simultaneously is an alternative scheme to improve energy efficiency and boost renewable energy. This paper proposes a phosphoric acid fuel cell trigeneration system integrated with methanol and steam reforming assisted by solar thermal energy. The trigeneration system consists of a solar heat collection subsystem, methanol steam reforming subsystem, fuel cell power generation subsystem, and recovered heat utilization subsystem. Their respective thermodynamic models are constructed to simulate the system input/output characteristics, and energy and exergy efficiencies are employed to evaluate the system thermodynamic performances. The contribution of solar energy to the system is analyzed using solar energy/exergy share. Through the simulation and analysis of methanol and steam reforming reactions, the optimal reaction pressure, temperature, and methanol to water ratio are obtained to improve the flow rate and content of produced hydrogen. The thermodynamic simulations of the trigeneration system show that the system energy efficiencies at the summer and winter design work conditions are 73.7% and 51.7%, while its exergy efficiencies are 18.8% and 26.1%, respectively. When the solar radiation intensity is different from the design work condition, the total energy and exergy efficiencies in winter decrease approximately by 4.7% and 2.2%, respectively, due to the decrease in solar heat collection efficiency. This proposed novel trigeneration system complemented by solar heat energy and methanol chemical energy is favorable for improving the

  14. Impact of fuel fabrication and fuel management technologies on uranium management

    International Nuclear Information System (INIS)

    Arnsberger, P.L.; Stucker, D.L.

    1994-01-01

    Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs

  15. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  16. DEVELOPING AN INTEGRATED NATIONAL STRATEGY FOR THE DISPOSITION OF SPENT NUCLEAR FUEL

    International Nuclear Information System (INIS)

    Gelles, C.M.

    2003-01-01

    This paper summarizes the Department of Energy's (DOE's) current efforts to strengthen its activities for the management and disposition of DOE-owned spent nuclear fuel (SNF). In August 2002 an integrated, ''corporate project'' was initiated by the Office of Environmental Management (EM) to develop a fully integrated strategy for disposition of the approximately ∼250,000 DOE SNF assemblies currently managed by EM. Through the course of preliminary design, the focus of this project rapidly evolved to become DOE-wide. It is supported by all DOE organizations involved in SNF management, and represents a marked change in the way DOE conducts its business. This paper provides an overview of the Corporate Project for Integrated/Risk-Driven Disposition of SNF (Corporate SNF Project), including a description of its purpose, scope and deliverables. It also summarizes the results of the integrated project team's (IPT's) conceptual design efforts, including the identification of project/system requirements and alternatives. Finally, this paper highlights the schedule of the corporate project, and its progress towards development of a DOE corporate strategy for SNF disposition

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-08-01

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

  18. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  19. Fuels planning: science synthesis and integration; environmental consequences fact sheet 08: Evaluating sedimentation risks associated with fuel management

    Science.gov (United States)

    William Elliot; Pete Robichaud

    2005-01-01

    This fact sheet describes the sources of sediment in upland forest watersheds in the context of fuel management activities. It presents the dominant forest soil erosion processes, and the principles behind the new sediment delivery interface developed to aid in erosion analysis of fuel management projects.

  20. Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 03: visualizing forest structure and fuels

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    The software described in this fact sheet provides managers with tools for visualizing forest and fuels information. Computer-based landscape simulations can help visualize stand and landscape conditions and the effects of different management treatments and fuel changes over time. These visualizations can assist forest planning by considering a range of management...

  1. Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 04: role of silviculture in fuel treatments

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    The principal goals of fuel treatments are to reduce fireline intensities, reduce the potential for crown fires, improve opportunities for successful fire suppression, and improve forest resilience to forest fires. This fact sheet discusses thinning, and surface fuel treatments, as well as challenges associated with those treatments.

  2. Fuels planning: science synthesis and integration; environmental consequences fact sheet 04: wildlife responses to fuels treatments: key considerations

    Science.gov (United States)

    David Pilliod

    2004-01-01

    Managers face a difficult task in predicting the effects of fuels treatments on wildlife populations, mostly because information on how animals respond to fuels treatments is scarce or does not exist. This paper discusses key considerations-aspects of an animal's ecology and available information-that, despite the scarcity of information, may make predictions of...

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

    International Nuclear Information System (INIS)

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

    1991-07-01

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

  4. Analyses for licensing of new fuel types at Paks NPP

    International Nuclear Information System (INIS)

    Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

    2003-01-01

    In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

  5. Long distance relationships : the secret for fuel cell success? fuel cell developers and integrators form trans-oceanic partnerships to crash through cultural barriers

    International Nuclear Information System (INIS)

    Horwitz, J.

    2009-01-01

    The varieties of viable fuel cell applications and widely varying regional market conditions have created global partnerships among entities with complementary attributes. Although it may appear that domestic liaisons among culturally similar players spawned from industry clusters should provide the clearest route to success in this industry, it is the intercontinental groupings which are demonstrating the most potential. This paper discussed the global fuel cell challenge and the vertical integration of multi-national partnerships. The paper also discussed the current global stationary market in perspective. Fuel cells require unique maintenance, support, and refueling including operator instruction and a new supply infrastructure. The paper addressed the fact that fuel cells represent a disruptive technology. A telecom backup status report was also presented. Other topics that were discussed included developing markets as well as specific examples of global organizations such as Canadian Ballard and Danish Dantherm Power and their fuel cell application solutions. It was concluded that after an inconsistent history, fuel cells have finally achieved viability in the real world. However, there is significant cultural resistance to their implementation in the United States. 4 figs

  6. Biomass gasification integrated with a solid oxide fuel cell and Stirling engine

    International Nuclear Information System (INIS)

    Rokni, Masoud

    2014-01-01

    An integrated gasification solid oxide fuel cell (SOFC) and Stirling engine for combined heat and power application is analyzed. The target for electricity production is 120 kW. Woodchips are used as gasification feedstock to produce syngas, which is then used to feed the SOFC stacks for electricity production. Unreacted hydrocarbons remaining after the SOFC are burned in a catalytic burner, and the hot off-gases from the burner are recovered in a Stirling engine for electricity and heat production. Domestic hot water is used as a heat sink for the Stirling engine. A complete balance-of-plant is designed and suggested. Thermodynamic analysis shows that a thermal efficiency of 42.4% based on the lower heating value (LHV) can be achieved if all input parameters are selected conservatively. Different parameter studies are performed to analyze the system behavior under different conditions. The analysis shows that the decreasing number of stacks from a design viewpoint, indicating that plant efficiency decreases but power production remains nearly unchanged. Furthermore, the analysis shows that there is an optimum value for the utilization factor of the SOFC for the suggested plant design with the suggested input parameters. This optimum value is approximately 65%, which is a rather modest value for SOFC. In addition, introducing a methanator increases plant efficiency slightly. If SOFC operating temperature decreases due to new technology then plant efficiency will slightly be increased. Decreasing gasifier temperature, which cannot be controlled, causes the plant efficiency to increase also. - Highlights: • Design of integrated gasification with solid oxide fuel and Stirling engine. • Important plant parameters study. • Plant running on biomass with and without methanator. • Thermodynamics of integrated gasification SOFC-Stirling engine plants

  7. A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

    International Nuclear Information System (INIS)

    Talamo, A.

    2009-01-01

    In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

  8. Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

  9. Integrated HT-PEMFC and multi-fuel reformer for micro CHP. Final report

    Energy Technology Data Exchange (ETDEWEB)

    2010-07-01

    reformed both methane and biogas although the efficiency was low, on the order of 15% due to excessive slip and heat losses. The construction and test of an integrated micro CHP system revealed several problems with the core technology. Therefore, rather than working with the integrated system, individual system components were tested separately. In spite of the problems with the reformer and the fuel cell stack the system was successfully operated and an electric efficiency of 18%{sub LHV} was demonstrated. (Author)

  10. Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

    International Nuclear Information System (INIS)

    Bracey, William; Bondre, Jayant; Shelton, Catherine; Edmonds, Robert

    2013-01-01

    The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In January 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to

  11. Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

    Energy Technology Data Exchange (ETDEWEB)

    Bracey, William; Bondre, Jayant; Shelton, Catherine [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States); Edmonds, Robert [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)

    2013-07-01

    The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In January 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to

  12. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  13. 16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

    International Nuclear Information System (INIS)

    Chambers, Martin; Kurincic, Bojan

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has experienced leaking fuel after increasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting. NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mixing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down...), operational cycle duration increase from 12 to 18 months (increasing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate). The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier. The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature

  14. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  15. Rural bioenergy: innovative integrated biosystem design for fuel and food from agrowastes

    Energy Technology Data Exchange (ETDEWEB)

    Selvam, Pagandai V. Pannir; Santiago, Brunno Henrique de S. [Universidade Federal de Alagoas (UFAL), Maceio, AL (Brazil)], email: pannirbr@gmail.com; Almeida, Louizy Minora C.A. de; Israel, Samy B.S. [Universidade Federal do Rio Grande do Norte (DEQ/CT/UFRN), Natal, RN (Brazil). Centro de Tecnologia. Dept. de Engenharia Quimica

    2008-07-01

    Brazil is the leader known for its ethanol biofuel development, but also for biomass charcoal, yet lacks in clean rural biofuel production. This paper deals with the system design for sustainable rural projects developments based on the the bioenergy production from biomass wastes using Innovative process equipment design and the process optimization. With the economic objective towards sustainable clean small scale rural energy production, and also the co-production of hot, cold thermal energy, at first, several preliminary projects had been developed and analysed. Then, a detailed project engineering and the results on economical viability were obtained. The dynamic material, energy and cash flow models of the complex integrated food processing and small scale energy productions, were at first constructed using excel electronic spread sheet for Windows. Computer aided system for process modeling and simulation using SUPERPRO, Intelligent Inc. Economic feasibilities studies using genetics algorithms are under developments. The preliminary analysis of cost and investments of the the integrated biomass utilisation projects which are included for this study are: biological aerobic treatment process, methane gas production using anaerobic digestion process, aerobic composting, drying of tropical fruits, effluent treatments, biogas and syngas production as well as cogeneration of energy production for drying and cooling using biogas. These models allowed us the identification of tech economical and scale up problems. The results obtained with several preliminary project development with few case studies are reported for integrated project developments for fuel and food using process and cost simulation models. Several economic problems related with implementation of the small scale rural energy system for sustained local developments based on tropical fruit producing rural areas are analysed and a concept of the integrated bio system for micro enterprise has been

  16. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

    1994-01-01

    A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

  17. Integrated production of wood fuel and pulp wood from young stands; Integroitujen tuotantomenetelmien vertailu

    Energy Technology Data Exchange (ETDEWEB)

    Korpilahti, A [Metsaeteho Oy, Helsinki (Finland)

    1997-12-01

    The aim of the study was to clarify the competitiveness of different harvesting chains and processing methods of first thinning wood. Great expectations have been laid on integrated production of wood fuel and pulp wood. Results produced in other bioenergy projects were taken into account, and in this project some field experiments on mechanised felling-bunching and compressing of the load of tree sections during forwarding were carried out. The new processing methods, the MASSAHAKE-method and chain-flail delimbing combined with small-scale drum debarking, still are under development giving a rather unstable data for comparisons. Both in pine and birch dominant stands modern multiple tree logging gave the most favourable results when ranking on the bases of the price of pulp chips. Integrated methods were not very far and they have more potential than methods based on harvesting delimbed short wood. When compared on the bases of the production cost of pulp, integrated methods were in general the most favourable because they give good subsidies on the form of bioenergy. (orig.)

  18. Development of multimedia computer-based training for VXI integrated fuel monitors

    International Nuclear Information System (INIS)

    Keeffe, R.; Ellacott, T.; Truong, Q.S.

    1999-01-01

    The Canadian Safeguards Support Program has developed the VXI Integrated Fuel Monitor (VFIM) which is based on the international VXI instrument bus standard. This equipment is a generic radiation monitor which can be used in an integrated mode where several detection systems can be connected to a common system where information is collected, displayed, and analyzed via a virtual control panel with the aid of computers, trackball and computer monitor. The equipment can also be used in an autonomous mode as a portable radiation monitor with a very low power consumption. The equipment has been described at previous international symposia. Integration of several monitoring systems (bundle counter, core discharge monitor, and yes/no monitor) has been carried out at Wolsong 2. Performance results from one of the monitoring systems which was installed at CANDU nuclear stations are discussed in a companion paper at this symposium. This paper describes the development of an effective multimedia computer-based training package for the primary users of the equipment; namely IAEA inspectors and technicians. (author)

  19. A microbial fuel cell–membrane bioreactor integrated system for cost-effective wastewater treatment

    International Nuclear Information System (INIS)

    Wang, Yong-Peng; Liu, Xian-Wei; Li, Wen-Wei; Li, Feng; Wang, Yun-Kun; Sheng, Guo-Ping; Zeng, Raymond J.; Yu, Han-Qing

    2012-01-01

    Highlights: ► An MFC–MBR integrated system for wastewater treatment and electricity generation. ► Stable electricity generation during 1000-h continuous operation. ► Low-cost electrode, separator and filter materials were adopted. -- Abstract: Microbial fuel cell (MFC) and membrane bioreactor (MBR) are both promising technologies for wastewater treatment, but both with limitations. In this study, a novel MFC–MBR integrated system, which combines the advantages of the individual systems, was proposed for simultaneous wastewater treatment and energy recovery. The system favored a better utilization of the oxygen in the aeration tank of MBR by the MFC biocathode, and enabled a high effluent quality. Continuous and stable electricity generation, with the average current of 1.9 ± 0.4 mA, was achieved over a long period of about 40 days. The maximum power density reached 6.0 W m −3 . Moreover, low-cost materials were used for the reactor construction. This integrated system shows great promise for practical wastewater treatment application.

  20. Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Badwan, Faris M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott F [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-06

    Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the