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Sample records for sodium cooled zirconium hydride moderated reactors

  1. METHOD OF FABRICATING A URANIUM-ZIRCONIUM HYDRIDE REACTOR CORE

    DOEpatents

    Weeks, I.F.; Goeddel, W.V.

    1960-03-22

    A method is described of evenly dispersing uranlum metal in a zirconium hydride moderator to produce a fuel element for nuclear reactors. According to the invention enriched uranium hydride and zirconium hydride powders of 200 mesh particle size are thoroughly admixed to form a mixture containing 0.1 to 3% by weight of U/sup 235/ hydride. The mixed powders are placed in a die and pressed at 100 tons per square inch at room temperature. The resultant compacts are heated in a vacuum to 300 deg C, whereby the uranium hydride deoomposes into uranium metal and hydrogen gas. The escaping hydrogen gas forms a porous matrix of zirconium hydride, with uramum metal evenly dispersed therethrough. The advantage of the invention is that the porosity and uranium distribution of the final fuel element can be more closely determined and controlled than was possible using prior methods of producing such fuel ele- ments.

  2. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    NASA Astrophysics Data System (ADS)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  3. METHOD OF MAKING DELTA ZIRCONIUM HYDRIDE MONOLITHIC MODERATOR PIECES

    DOEpatents

    Vetrano, J.B.

    1962-01-23

    A method is given for preparing large, sound bodies of delta zirconium hydride. The method includes the steps of heating a zirconium body to a temperature of not less than l000 deg C, providing a hydrogen atmosphere for the zirconium body at a pressure not greater than one atmosphere, reducing the temperature slowly to 800 deg C at such a rate that cracks do not form while maintaining the hydrogen pressure substantially constant, and cooling in an atmosphere of hydrogen. (AEC)

  4. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  5. Methods of mathematical statistics for verification of hydrogen content in zirconium hydride moderator

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Glushkov, Ye.S.; Kompaniets, G.V.; Nosov, V.I. )

    1995-02-01

    The hydrogen content of zirconium hydride blocks used as the moderator in Topaz-2-type space reactors is estimated according to correlation-regression analysis procedures of mathematical statistics and is based on the results of the definition of the reactivity of the blocks in a research critical assembly. A linear mathematical model for a variable response is formulated within the framework of the first-order perturbation theory applied to the estimation of reactivity effects in reactors. A PASPORT computer code is written based on the developed algorithm. The statistical analysis of the available data performed by using PASPORT shows that the developed approach allows determination of the insignificance of the contribution of the impurities to the reactivity of the blocks, verification of the manufacturer's data on the hydrogen content in zirconium hydride blocks, and estimation of the reactivity shift in a standard block.

  6. Space nuclear power system based on thermionic reactor with single-cell TFEs and zirconium hydride moderator

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A. ); Nickitin, V.P.; Ogloblin, B.G.; Lutov, J.I.; Luppov, A.N.; Gabrusev, V.N.; Klimov, A.V. ); Nicolaev, J.V.; Kucherov, R.J.; Eremin, S.A. )

    1993-01-15

    The results of research and development work on the space nuclear power system with net electric power of 40 kW performed with cooperation of Russian research and engineering organizations in 1991--92 are presented in this paper. Development work was carried out on the basis of experience gained in the course of Topaz-2 SNPS creation. As the basis for SNPS the 127-channel intermediate reactor is used with in-core single-cell TFEs having improved parameters. The SNPS parameters, distinctive features of decision on design and layout, safety ensuring concepts, proposals on the future stages of SNPS development are also considered in the paper.

  7. The Industrial Sodium Cooled Fast Reactor

    SciTech Connect

    Samuel E. Bays; Haihua Zhao; Hongbin Zhang

    2009-04-01

    This paper investigates the use of enrichment and moderator zoning methods for optimizing the r-z power distribution within sodium cooled fast reactors. These methods allow overall greater fuel utilization in the core resulting in more fuel being irradiated near the maximum allowed thermal power. The peak-to-average power density was held to 1.18. This core design, in conjunction with a multiple-reheat Brayton power conversion system, has merit for producing an industrial level of electrical output (2400MWth, 1000MWe) from a relatively compact core size. The total core radius, including reflectors and shields, was held to 1.78m. Preliminary safety analysis suggests that positive reactivity insertion resulting from a leak between the sodium primary loop and helium power conversion system can be mitigated using simple gas-liquid centripetal separation strategies in the plant’s primary loop.

  8. Bulk Hydrides and Delayed Hydride Cracking in Zirconium Alloys

    NASA Astrophysics Data System (ADS)

    Tulk, Eric F.

    Zirconium alloys are susceptible to engineering problems associated with the uptake of hydrogen throughout their design lifetime in nuclear reactors. Understanding of hydrogen embrittlement associated with the precipitation of brittle hydride phases and a sub-critical crack growth mechanism known as Delayed Hydride Cracking (DHC) is required to provide the engineering justifications for safe reactor operation. The nature of bulk zirconium hydrides at low concentrations (< 100 wt. ppm) is subject to several contradictory descriptions in the literature associated with the stability and metastability of gamma-phase zirconium hydride. Due to the differing volume expansions (12-17%) and crystallography between gamma and delta hydride phases, it is suggested that the matrix yield strength may have an effect on the phase stability. The present work indicated that although yield strength can shift the phase stability, other factors such as microstructure and phase distribution can be as or more important. This suggests that small material differences are the reason for the literature discrepancies. DHC is characterised by the repeated precipitation, growth, fracture of brittle hydride phases and subsequent crack arrest in the ductile metal. DHC growth is associated primarily the ability of hydrogen to diffuse under a stress induced chemical potential towards a stress raiser. Knowledge of the factors controlling DHC are paramount in being able to appropriately describe DHC for engineering purposes. Most studies characterise DHC upon cooling to the test temperature. DHC upon heating has not been extensively studied and the mechanism by which it occurs is somewhat controversial in the literature. This work shows that previous thermo-mechanical processing of hydrided zirconium can have a significant effect on the dissolution behaviour of the bulk hydride upon heating. DHC tests with gamma-quenched, furnace cooled-delta and reoriented bulk hydrides upon heating and DHC upon

  9. A resting bottom sodium cooled fast reactor

    SciTech Connect

    Costes, D.

    2012-07-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  10. Superstoichiometric hydride of zirconium

    SciTech Connect

    Kupryazhkin, A.Ya.; Shchepetkin, A.A.; Zabolotskaya, E.V.; Pletnev, R.N.; Alyamovskii, S.I.; Kitaev, G.A.

    1987-12-01

    Superstoichiometric hydrides of zirconium have been obtained all the way up to the composition ZrH/sub 2.4/ by additional hydrogenation of ZrH/sub 2/ as a result of redistribution of hydrogen atoms between t- and o-positions. In the preparation of the hydrides the authors used zirconium iodide with an impurity content no greater than 10/sup -2/ to 10/sup -2/ mole %; the hydrogen and helium used in this work had a minimum purity of 99.95%. The content of hydrogen in the specimens was determined by a volumetric method. The x-ray diffraction analysis was performed in a DRON-2.0 unit (CuK/sub ..cap alpha../ radiation). PMR spectra were recorded in a broad-line spectrometer in the temperature interval 150-450 K.

  11. Process for massively hydriding zirconium--uranium fuel elements

    DOEpatents

    Katz, N.H.

    1973-12-01

    A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

  12. MORSE Monte Carlo shielding calculations for the zirconium hydride reference reactor

    NASA Technical Reports Server (NTRS)

    Burgart, C. E.

    1972-01-01

    Verification of DOT-SPACETRAN transport calculations of a lithium hydride and tungsten shield for a SNAP reactor was performed using the MORSE (Monte Carlo) code. Transport of both neutrons and gamma rays was considered. Importance sampling was utilized in the MORSE calculations. Several quantities internal to the shield, as well as dose at several points outside of the configuration, were in satisfactory agreement with the DOT calculations of the same.

  13. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  14. Shape optimization of a sodium cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  15. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    SciTech Connect

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  16. Energy Distributions of Neutrons Scattered from Graphite, Light and Heavy Water, Ice, Zirconium Hydride, Lithium Hydride, Sodium Hydride and Ammonium Chloride by the Beryllium Detector Method

    DOE R&D Accomplishments Database

    Woods, A. D. B.; Brockhouse, Bertram N.; Sakamoto, M.; Sinclair, R. N.

    1960-09-12

    Energy distributions of neutrons scattered from various moderators and from several hydrogenous substances were measured at energy transfers of 0.02 to 0.24 ev. Results from experiments on graphite, light and heavy water, ice, ZrH, LiH, NaH, and NH4Cl are included. It is noted that the results are of a preliminary character; however, they are probably the most accurate measurements of high-energy transfers yet made. (J.R.D.)

  17. Method of making crack-free zirconium hydride

    DOEpatents

    Sullivan, Richard W.

    1980-01-01

    Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

  18. Fuels for sodium-cooled fast reactors: US perspective

    NASA Astrophysics Data System (ADS)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.

    2007-09-01

    The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50 000 MOX rods, over 130 000 metal rods, and 600 mixed carbide rods, in EBR-II and FFTF alone. All three types have been demonstrated capable of fuel utilization at or above 200 GWd/MTHM. To varying degrees, life-limiting phenomena for each type have been identified and investigated, and there are no disqualifying safety-related fuel behaviors. All three fuel types appear capable of meeting requirements of sodium-cooled fast reactor fuels, with reliability of mixed oxide and metal fuel well established. Improvements in irradiation performance of cladding and duct alloys have been a key development in moving these fuel designs toward higher-burnup potential. Selection of one fuel system over another will depend on circumstances particular to the application and on issues other than fuel performance, such as fabrication cost or overall system safety performance.

  19. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    SciTech Connect

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  20. Sodium leak detection system for liquid metal cooled nuclear reactors

    DOEpatents

    Modarres, Dariush

    1991-01-01

    A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

  1. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  2. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  3. Simulation of ultrasonic inspection for sodium cooled reactors using CIVA

    SciTech Connect

    Reverdy, F.; Baque, F.; Lu, B.; Jezzine, K.; Dorval, V.; Augem, J. M.

    2011-07-01

    In-service inspection of sodium fast reactors (SFR) requires the development of non-destructive techniques adapted to the harsh conditions of the environment (opaque and hot) and the complexity of the examination (large and littered reactor block). Ultrasonic techniques are seen as suitable candidates for the inspection of SFRs and two approaches are being followed: inside inspection where transducers are directly immersed in sodium coolant and inspection from outside with transducers positioned along the wall of the main vessel. Probe design and inspection performances can be predicted by using comprehensive models that can take into account the various variables of the problem. These models are explained in this paper. (authors)

  4. Zirconium hydride containing explosive composition

    DOEpatents

    Walker, Franklin E.; Wasley, Richard J.

    1981-01-01

    An improved explosive composition is disclosed and comprises a major portion of an explosive having a detonation velocity between about 1500 and 10,000 meters per second and a minor amount of a donor additive comprising a non-explosive compound or mixture of non-explosive compounds which when subjected to an energy fluence of 1000 calories/cm.sup.2 or less is capable of releasing free radicals each having a molecular weight between 1 and 120. Exemplary donor additives are dibasic acids, polyamines and metal hydrides.

  5. Ways of improvement for the materials of sodium cooled fast reactors

    SciTech Connect

    Horowitz, E.

    2012-07-01

    The French sodium cooled prototype reactor ASTRID will take into account 'Generation IV' requirements, especially a long operational life-time (60 years) and a high efficiency. The good behavior of austenitic steel AISI316L(N), should be confirmed for a use, in moderately irradiated and unirradiated parts of ASTRID. Parts recovered from dismantled French sodium-cooled reactors will be characterized. Further experiments must be carried out concerning ageing of these components. Other materials will be chosen for fuel wrapping and cladding, in order to reduce creep and swelling under irradiation, (either conventional, or oxide-dispersed strengthened steels (ODSS). Corrosion of ODSS in the presence of sodium needs a serious assessment The lifetime of primary pumps components made of Duplex steels should also be assessed. The disruptions in steam generator tubes should be minimized and controlled; therefore, optimised designs and geometries must be established before defining the corresponding materials. Either Modified 9Cr1Mo or Incoloy 800H, might be candidates;it will be necessary to check whether austenitic steels are compatible with Modified 9Cr1Mo or Incoloy 800H in the same circuit. For all materials, the best manufacturing processes must be combined with thermal, mechanical treatments; calculations of phase diagrams (CALPHAD) might be used to optimise both treatments and chemical compositions. (authors)

  6. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  7. Status of highly loaded uranium-zirconium hydride Low Enriched Uranium (LEU) fuel programs

    SciTech Connect

    West, G.B.

    1984-07-01

    The LEU uranium-zirconium hydride fuel has successfully completed all essential development testing and is offered commercially in both the 20 weight percent and 45 weight percent uranium concentrations for existing TRIGA reactors and as a conversion fuel for plate-type reactors. Several reactor facilities have already begun conversion to LEU uranium-zirconium hydride fuel. Currently, U-ZrH LEU fuel is in use in a mixed plate and rod fuel configuration in the 1.0 MW reactor at the National Tsing Hua University in Taipei, Taiwan. A complete core of 20 wt-$ LEU fuel has been fabricated for the 3 MW forced flow TRIGA Mark II reactor system with pulsing capability scheduled to be completed in 1984 in Bangladesh. Also, a complete core of 20 vt-% 4-rod shrouded cluster fuel has been fabricated and shipped for conversion and upgrading of the PRR-1 reactor in the Philippines. The converted reactor will operate at 3 MW with forced cooling and have pulsing capability. Startup is now scheduled for early 1985. U-ZrH LEU fuel has also been delivered to existing TRIGA reactor facilities in Malaysia, Thailand and Yugoslavia and is being put to use as new fuel is needed to meet burnup requirements. For stepwise conversions of higher power reactors (>5 MW), where flow distribution and neutron spectrum effects are of greater importance, a detailed analysis and study is a prerequisite. A joint program involving GA, ANL, and the Cekmece Center is now in progress for analysis of a single U-ZrH LEU fuel test cluster operating in the 5 MW HEU plate-type core of the TR-2 reactor at the Cekmece Nuclear Research and Training Center in Istanbul, Turkey. Inserting this one cluster would be the initial step in a stepwise conversion of the HEU plate-type core to U-ZrH LEU fuel.

  8. SODIUM DEUTERIUM REACTOR

    DOEpatents

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  9. Modelling zirconium hydrides using the special quasirandom structure approach.

    PubMed

    Wang, H; Chroneos, A; Jiang, C; Schwingenschlögl, U

    2013-05-28

    The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH(2-x) (x = 0.25-0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys. PMID:23588997

  10. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  11. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-20

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  12. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    SciTech Connect

    Haihua Zhao; Hongbin Zhang

    2007-11-01

    The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

  13. A quantitative phase field model for hydride precipitation in zirconium alloys: Part II. Modeling of temperature dependent hydride precipitation

    NASA Astrophysics Data System (ADS)

    Xiao, Zhihua; Hao, Mingjun; Guo, Xianghua; Tang, Guoyi; Shi, San-Qiang

    2015-04-01

    A quantitative free energy functional developed in Part I (Shi and Xiao, 2014 [1]) was applied to model temperature dependent δ-hydride precipitation in zirconium in real time and real length scale. At first, the effect of external tensile load on reorientation of δ-hydrides was calibrated against experimental observations, which provides a modification factor for the strain energy in free energy formulation. Then, two types of temperature-related problems were investigated. In the first type, the effect of temperature transient was studied by cooling the Zr-H system at different cooling rates from high temperature while an external tensile stress was maintained. At the end of temperature transients, the average hydride size as a function of cooling rate was compared to experimental data. In the second type, the effect of temperature gradients was studied in a one or two dimensional temperature field. Different boundary conditions were applied. The results show that the hydride precipitation concentrated in low temperature regions and that it eventually led to the formation of hydride blisters in zirconium. A brief discussion on how to implement the hysteresis of hydrogen solid solubility on hydride precipitation and dissolution in the developed phase field scheme is also presented.

  14. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  15. METHOD OF PREPARING SINTERED ZIRCONIUM METAL FROM ITS HYDRIDES

    DOEpatents

    Angier, R.P.

    1958-02-11

    The invention relates to the preparation of metal shapes from zirconium hydride by powder metallurgical techniques. The zirconium hydride powder which is to be used for this purpose can be prepared by rendering massive pieces of crystal bar zirconium friable by heat treatment in purified hydrogen. This any then be ground into powder and powder can be handled in the air without danger of it igniting. It may then be compacted in the normal manner by being piaced in a die. The compact is sintered under vacuum conditions preferably at a temperature ranging from 1200 to 1300 deg C and for periods of one to three hours.

  16. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  17. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, A.B. Jr.; Levy, I.S.; Trimble, D.J.; Lanning, D.D.; Gerber, F.S.

    1990-04-10

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-based materials is disclosed. Samples of zirconium-based materials having different compositions and/or fabrication methods are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280 to 316 C). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hydriding for the same materials when subject to in-reactor (irradiated) corrosion. 1 figure.

  18. Current Design Status of Sodium Cooled Super-Safe,Small and Simple Reactor

    SciTech Connect

    Ueda, N.; Knoshita, I.; Nishi, Y.; Minato, A.; Yokoyama, T.; Nishiguchi, Y.

    2002-07-01

    CRIEPI has been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. And a conceptual design of 4S (Super-Safe, Small and Simple) reactor was proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void coefficient are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core life time is more than 10 years; (5) Its construction, maintenance and operation are expected to be very simple by eliminating active components from inside of a reactor vessel. The 4S reactor is a sodium cooled fast reactor and its reactivity is not controlled by neutron absorber rods but by neutron reflectors. An electrical output is 50 MW. This paper describes a design modification to enhance the feasibility from the previous 4S design. A core active height can be shortened to 1.5 m from 4.0 m to keep the reactivity characteristics. An averaged fuel burn-up is up to 70 GWD/ton and a pressure drop at the core region is less than 0.1 MPa. A reactivity control system is modified according with the core design change. As for the steam generator design, sodium-water reaction accidents must be taken into account as a design basis event for the utilization of the secondary sodium coolant. Therefore, a modified plate type heat exchanger is proposed as a steam generator. It may be possible to develop a compact steam generator, which is free from sodium-water reaction accidents and to eliminate the secondary sodium systems. The 4S reactor without secondary system has been proposed as a candidate design. (authors)

  19. A laboratory method to predict hydriding properties of zirconium alloys under irradiation

    SciTech Connect

    Johnson, A.B. Jr.; Levy, I.S.; Lanning, D.D.; Gerber, F.S.; Trimble, D.J.

    1988-08-01

    A corrosion and hydriding test series on zirconium alloys in the Engineering Test Reactor G-7 loop demonstrated relatively large lot-to-lot and alloy-to-alloy differences in hydriding rates under irradiation. Similar differences were also found among irradiated Zircaloy-2 pressure tubes fabricated by three suppliers for the Hanford Site N Reactor. This substantial in-reactor hydriding data base and access to archive materials from these alloys permitted an investigation of methods to reproduce the in-reactor hydriding orders-of-merit by an out-of-reactor method. The out-of-reactor method selected for investigation consisted of autoclaving alloys in relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solutions. The test times ranged from 7 to 35 d, and the samples were held at constant temperatures within the water reactor coolant temperature range (280/degree/C to 315/degree/C). The in-reactor hydriding behavior for several lots of Zircaloy-2, one lot of Zircaloy-4, and one lot of Zr--2.5Nb was reproduced in the lithium hydroxide tests. The hydriding rates were compared on the basis of the ratio of hydrogen weight gain to oxide weight gain. 13 refs., 4 figs., 6 tabs.

  20. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K.

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  1. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  2. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  3. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  4. The effect of stress state on zirconium hydride reorientation

    NASA Astrophysics Data System (ADS)

    Cinbiz, Mahmut Nedim

    Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by

  5. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  6. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  7. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, Jr., A. Burtron; Levy, Ira S.; Trimble, Dennis J.; Lanning, Donald D.; Gerber, Franna S.

    1990-01-01

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

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

    SciTech Connect

    Sindelar, R.; Louthan, M.; PNNL, B.

    2015-05-29

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

  9. Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

    SciTech Connect

    Zhang, Hongbin; Zhao, Haihua; Davis, Cliff

    2008-09-01

    An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems.

  10. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  11. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the

  12. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  13. Concept for Dismantling the Reactor Vessel and the Biological Shield of the Compact Sodium-Cooled Nuclear Reactor Facility (KNK)

    SciTech Connect

    Hillebrand, I.; Benkert, J.

    2002-02-27

    The Compact Sodium-cooled Nuclear Reactor Facility (KNK) was an experimental nuclear power plant of 20 MW electric power erected on the premises of the Karlsruhe Research Center. The plant was initially run as KNK I with a thermal core between 1971 and 1974 and then, between 1977 and 1991, with a fast core as the KNK II fast breeder plant. Under the decommissioning concept, the plant is to be decommissioned completely to green field conditions at the end of 2005 in ten steps, i.e. under the corresponding ten decommissioning permits. To this day, nine decommissioning permits have been issued, the first one in 1993 and the most recent one, number nine, in 2001. The decommissioning and demolition activities covered by decommissioning permits 1 to 7 have been completed. Under the 8th Decommissioning Permit, the components of the primary system and the rotating reactor top shield are to be removed by late 2001. Under the 9th Decommissioning Permit, the reactor vessel with its internals, the primary shield, and the biological shield are to be dismantled. The residual sodium volume in the reactor vessel was estimated to amount to approx. 30 l. The maximum Co-60 activation is on the order of 107-108 Bq/g; the maximum dose rate in the middle of the vessel was measured in April 1997 to be 55 Sv/h. The difficulty involved especially in dismantling KNK, on the one hand, is posed by the residual sodium in the plant, which determines the choice of neither wet nor thermical techniques to be used in disassembly. Another difficulty is caused by the depth of activation by fast neutrons, as a result of which not only the reactor vessel proper, but also the entire primary shield (60 cm of grey cast iron) and large parts of the biological shield must be disassembled and disposed of under remote control.

  14. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

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

    SciTech Connect

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

    1994-03-01

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

  16. Impact of nuclear data on sodium-cooled fast reactor calculations

    NASA Astrophysics Data System (ADS)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  17. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect

    Yamano, H.; Kubo, S.; Tani, A.; Nishino, H.; Sakai, T.

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  18. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  19. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    SciTech Connect

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2016-01-01

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  20. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect

    Kim, T. K.; Grandy, C.; Hill, R. N.

    2012-07-01

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

  1. Conceptual design features of the Kalimer-600 sodium cooled fast reactor

    SciTech Connect

    Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum; Jeong, Hae-Yong

    2007-07-01

    An advanced sodium cooled fast reactor concept, KALIMER-600, has been developed by the Korea Atomic Energy Research Institute to satisfy the Gen-IV technology goals of sustainability, safety and reliability, economics and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on a proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design were verified through a safety analysis of its bounding events. The results for various unprotected events imply that the KALIMER-600 design can accommodate all the analyzed ATWS events. This self-regulation capability of the power without a scram is mainly attributed to the inherent reactivity feedback mechanisms implemented in the metal fuel core design and completely passive decay heat removal system. (authors)

  2. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  3. A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel

    SciTech Connect

    Samuel E. Bays

    2011-02-01

    An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

  4. Mechanical properties of zirconium alloys and zirconium hydrides predicted from density functional perturbation theory

    SciTech Connect

    Weck, Philippe F.; Kim, Eunja; Tikare, Veena; Mitchell, John A.

    2015-10-13

    Here, the elastic properties and mechanical stability of zirconium alloys and zirconium hydrides have been investigated within the framework of density functional perturbation theory. Results show that the lowest-energy cubic Pn-3m with combining macron]m polymorph of δ-ZrH1.5 does not satisfy all the Born requirements for mechanical stability, unlike its nearly degenerate tetragonal P42/mcm polymorph. Elastic moduli predicted with the Voigt–Reuss–Hill approximations suggest that mechanical stability of α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates is limited by the shear modulus. According to both Pugh's and Poisson's ratios, α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates can be considered ductile. The Debye temperatures predicted for γ-ZrH, δ-ZrH1.5 and ε-ZrH2 are θD = 299.7, 415.6 and 356.9 K, respectively, while θD = 273.6, 284.2, 264.1 and 257.1 K for the α-Zr, Zry-4, ZIRLO and M5 matrices, i.e. suggesting that Zry-4 possesses the highest micro-hardness among Zr matrices.

  5. Mechanical properties of zirconium alloys and zirconium hydrides predicted from density functional perturbation theory

    DOE PAGESBeta

    Weck, Philippe F.; Kim, Eunja; Tikare, Veena; Mitchell, John A.

    2015-10-13

    Here, the elastic properties and mechanical stability of zirconium alloys and zirconium hydrides have been investigated within the framework of density functional perturbation theory. Results show that the lowest-energy cubic Pn-3m with combining macron]m polymorph of δ-ZrH1.5 does not satisfy all the Born requirements for mechanical stability, unlike its nearly degenerate tetragonal P42/mcm polymorph. Elastic moduli predicted with the Voigt–Reuss–Hill approximations suggest that mechanical stability of α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates is limited by the shear modulus. According to both Pugh's and Poisson's ratios, α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates can be considered ductile. The Debye temperatures predicted formore » γ-ZrH, δ-ZrH1.5 and ε-ZrH2 are θD = 299.7, 415.6 and 356.9 K, respectively, while θD = 273.6, 284.2, 264.1 and 257.1 K for the α-Zr, Zry-4, ZIRLO and M5 matrices, i.e. suggesting that Zry-4 possesses the highest micro-hardness among Zr matrices.« less

  6. Hydriding process

    DOEpatents

    Raymond, J.W.; Taketani, H.

    1973-12-01

    BS>A method is described for hydriding a body of a Group IV-B metal, preferably zirconium, to produce a crack-free metal-hydride bedy of high hydrogen content by cooling the body at the beta to beta + delta boundary, without further addition of hydrogen, to precipitate a fine-grained delta-phase metal hydride in the beta + delta phase region and then resuming the hydriding, preferably preceded by a reheating step. (Official Gazette)

  7. Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

    SciTech Connect

    Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

    2006-07-01

    The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

  8. ASTRID sodium cooled fast reactor: Program for improving in service inspection and repair

    SciTech Connect

    Jadot, F.; De Dinechin, G.; Augem, J. M.; Sibilo, J.

    2011-07-01

    In the frame of the CEA, EDF, AREVA coordinated research program for the development of Generation IV sodium-cooled fast reactors (SFR), the ASTRID project was launched in 2010. For the future prototype, the improvement of in-service inspection and repair (ISI and R) capabilities was identified as a major issue. Following the pluri-annual SFR research program, the ISI and R main R and D axes remain: i) improvement of the primary system conceptual design, ii) development of measurement and inspection techniques (continuous monitoring instrumentation and periodic inspection tools), iii) accessibility and associated robotics, and iv) development and validation of repair processes. Associated ISI and R needs are being defined through an iterative method between designers and instrumentation specialists: adaptation of the Design to ISI and R requirements, fission chamber development, validation of the ultrasonic and chemical transducers, of ultrasonic non destructive simulation, of acoustic surveillance, of laser repair intervention processes, of connected robotic equipment. Moreover, CEA, as leader of the ASTRID Project, is willing to find new contributors, partners or suppliers, in order to get innovative, diversified, exhaustive and efficient solutions. (authors)

  9. Effect of thermo-mechanical cycling on zirconium hydride reorientation studied in situ with synchrotron X-ray diffraction

    NASA Astrophysics Data System (ADS)

    Colas, Kimberly B.; Motta, Arthur T.; Daymond, Mark R.; Almer, Jonathan D.

    2013-09-01

    The circumferential hydrides normally present in nuclear reactor fuel cladding after reactor exposure may dissolve during drying for dry storage and re-precipitate when cooled under load into a more radial orientation, which could embrittle the fuel cladding. It is necessary to study the rates and conditions under which hydride reorientation may happen in order to assess fuel integrity in dry storage. The objective of this work is to study the effect of applied stress and thermal cycling on the hydride morphology in cold-worked stress-relieved Zircaloy-4 by combining conventional metallography and in situ X-ray diffraction techniques. Metallography is used to study the evolution of hydride morphology after several thermo-mechanical cycles. In situ X-ray diffraction performed at the Advanced Photon Source synchrotron provides real-time information on the process of hydride dissolution and precipitation under stress during several thermal cycles. The detailed study of diffracted intensity, peak position and full-width at half-maximum provides information on precipitation kinetics, elastic strains and other characteristics of the hydride precipitation process. The results show that thermo-mechanical cycling significantly increases the radial hydride fraction as well as the hydride length and connectivity. The radial hydrides are observed to precipitate at a lower temperature than circumferential hydrides. Variations in the magnitude and range of hydride strains due to reorientation and cycling have also been observed. These results are discussed in light of existing models and experiments on hydride reorientation. The study of hydride elastic strains during precipitation shows marked differences between circumferential and radial hydrides, which can be used to investigate the reorientation process. Cycling under stress above the threshold stress for reorientation drastically increases both the reoriented hydride fraction and the hydride size. The reoriented hydride

  10. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  11. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  12. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  13. Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor

    SciTech Connect

    Nashine, B.K.; Dash, S.K.; Gurumurthy, K.; Rajan, M.; Vaidyanathan, G.

    2006-07-01

    DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560 deg. C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at {approx} 560 deg. C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m{sup 3}/h capacity and developing 1.45 Kg/ cm{sup 2} pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel

  14. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  15. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  16. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    SciTech Connect

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-31

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  17. In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

    NASA Astrophysics Data System (ADS)

    Filliatre, P.; Jammes, C.; Chapoutier, N.; Jeannot, J.-P.; Jadot, F.; Batail, R.; Verrier, D.

    2014-04-01

    The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the "clean sodium" concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

  18. Micro-scale fracture experiments on zirconium hydrides and phase boundaries

    NASA Astrophysics Data System (ADS)

    Chan, H.; Roberts, S. G.; Gong, J.

    2016-07-01

    Fracture properties of micro-scale zirconium hydrides and phase boundaries were studied using microcantilever testing methods. FIB-machined microcantilevers were milled on cross-sectional surfaces of hydrided samples, with the most highly-stressed regions within the δ-hydride film, within the α-Zr or along the Zr-hydride interface. Cantilevers were notched using the FIB and then tested in bending using a nanoindenter. Load-displacement results show that three types of cantilevers have distinct deformation properties. Zr cantilevers deformed plastically. Hydride cantilevers fractured after a small amount of plastic flow; the fracture toughness of the δ-hydride was found to be 3.3 ± 0.4 MPam1/2 and SEM examination showed transgranular cleavage on the fracture surfaces. Cantilevers notched at the Zr-hydride interface developed interfacial voids during loading, at loads considerably lower than that which initiate brittle fracture of hydrides.

  19. Safe and Effective Deactivation of Metallic Sodium Filled Scrap and Cold Traps From Sodium-cooled Nuclear Reactor D and D - 12176

    SciTech Connect

    Nester, Dean; Crocker, Ben; Smart, Bill

    2012-07-01

    As part of the Plateau Remediation Project at US Department of Energy's Hanford, Washington site, CH2M Hill Plateau Remediation Company (CHPRC) contracted with IMPACT Services, LLC to receive and deactivate approximately 28 cubic meters of sodium metal contaminated debris from two sodium-cooled research reactors (Enrico Fermi Unit 1 and the Fast Flux Test Facility) which had been stored at Hanford for over 25 years. CHPRC found an off-site team composed of IMPACT Services and Commodore Advanced Sciences, Inc., with the facilities and technological capabilities to safely and effectively perform deactivation of this sodium metal contaminated debris. IMPACT Services provided the licensed fixed facility and the logistical support required to receive, store, and manage the waste materials before treatment, and the characterization, manifesting, and return shipping of the cleaned material after treatment. They also provided a recycle outlet for the liquid sodium hydroxide byproduct resulting from removal of the sodium from reactor parts. Commodore Advanced Sciences, Inc. mobilized their patented AMANDA unit to the IMPACT Services site and operated the unit to perform the sodium removal process. Approximately 816 Kg of metallic sodium were removed and converted to sodium hydroxide, and the project was accomplished in 107 days, from receipt of the first shipment at the IMPACT Services facility to the last outgoing shipment of deactivated scrap metal. There were no safety incidents of any kind during the performance of this project. The AMANDA process has been demonstrated in this project to be both safe and effective for deactivation of sodium and NaK. It has also been used in other venues to treat other highly reactive alkali metals, such as lithium (Li), potassium (K), NaK and Cesium (Cs). (authors)

  20. Measurement and modeling of strain fields in zirconium hydrides precipitated at a stress concentration

    SciTech Connect

    Allen, Gregory B.; Kerr, Matthew; Daymond, Mark R.

    2012-10-23

    Hydrogen adsorption into zirconium, as a result of corrosion in aqueous environments, leads to the precipitation of a secondary brittle hydride phase. These hydrides tend to first form at stress concentrations such as fretting flaws or cracks in engineering components, potentially degrading the structural integrity of the component. One mechanism for component failure is a slow crack growth mechanism known as Delayed Hydride Cracking (DHC), where hydride fracture occurs followed by crack arrest in the ductile zirconium matrix. The current work employs both an experimental and a modeling approach to better characterize the effects and behavior of hydride precipitation at such stress concentrations. Strains around stress concentrations containing hydrides were mapped using High Energy X-ray Diffraction (HEXRD). These studies highlighted important differences in the behavior of the hydride phase and the surrounding zirconium matrix, as well as the strain associated with the precipitation of the hydride. A finite element model was also developed and compared to the X-ray strain mapping results. This model provided greater insight into details that could not be obtained directly from the experimental approaches, as well as providing a framework for future modeling to predict the effects of hydride precipitation under varied conditions.

  1. Measurement and modeling of strain fields in zirconium hydrides precipitated at a stress concentration

    NASA Astrophysics Data System (ADS)

    Allen, Gregory B.; Kerr, Matthew; Daymond, Mark R.

    2012-11-01

    Hydrogen adsorption into zirconium, as a result of corrosion in aqueous environments, leads to the precipitation of a secondary brittle hydride phase. These hydrides tend to first form at stress concentrations such as fretting flaws or cracks in engineering components, potentially degrading the structural integrity of the component. One mechanism for component failure is a slow crack growth mechanism known as Delayed Hydride Cracking (DHC), where hydride fracture occurs followed by crack arrest in the ductile zirconium matrix. The current work employs both an experimental and a modeling approach to better characterize the effects and behavior of hydride precipitation at such stress concentrations. Strains around stress concentrations containing hydrides were mapped using High Energy X-ray Diffraction (HEXRD). These studies highlighted important differences in the behavior of the hydride phase and the surrounding zirconium matrix, as well as the strain associated with the precipitation of the hydride. A finite element model was also developed and compared to the X-ray strain mapping results. This model provided greater insight into details that could not be obtained directly from the experimental approaches, as well as providing a framework for future modeling to predict the effects of hydride precipitation under varied conditions.

  2. Performance of low smeared density sodium-cooled fast reactor metal fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  3. Performance of low smeared density sodium-cooled fast reactor metal fuel

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  4. Performance of Low Smeared Density Sodium-cooled Fast Reactor Metal Fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  5. AIR COOLED NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  6. Application of GRS method to evaluation of uncertainties of calculation parameters of perspective sodium-cooled fast reactor

    SciTech Connect

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A.

    2012-07-01

    A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used for estimation of errors of calculation parameters (K{sub eff}, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK. (authors)

  7. Performance of low smeared density sodium-cooled fast reactor metal fuel

    DOE PAGESBeta

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

  8. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    SciTech Connect

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  9. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    SciTech Connect

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  10. REACTOR MODERATOR STRUCTURE

    DOEpatents

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  11. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    SciTech Connect

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  12. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    DOEpatents

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  13. Method of pair exchange of fuel assemblies and its use in optimizing the energy distribution of water-cooled/water-moderated reactors

    SciTech Connect

    Simonov, V.D.; Pavlov, V.I.; Perminov, A.A.; Pechikin, V.A.; Filimonov, P.E.; Yuskov, A.M.

    1987-03-01

    The authors review various computer codes for determining the power distribution and optimizing the fueling procedure for water cooled and moderated reactors and assess their relative efficiencies in terms of computation time required. The algorithms take into account reactivity coefficients and neutron diffusion theory. A sensitivity analysis of the codes is given and steps are outlined for implementation of the codes for various reactor core configurations.

  14. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

    SciTech Connect

    Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

    2013-01-01

    Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for the advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.

  15. HEAVY WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  16. Hydrogen storage in sodium aluminum hydride.

    SciTech Connect

    Ozolins, Vidvuds; Herberg, J.L. (Lawrence Livermore National Laboratories, Livermore, CA); McCarty, Kevin F.; Maxwell, Robert S. (Lawrence Livermore National Laboratories, Livermore, CA); Stumpf, Roland Rudolph; Majzoub, Eric H.

    2005-11-01

    Sodium aluminum hydride, NaAlH{sub 4}, has been studied for use as a hydrogen storage material. The effect of Ti, as a few mol. % dopant in the system to increase kinetics of hydrogen sorption, is studied with respect to changes in lattice structure of the crystal. No Ti substitution is found in the crystal lattice. Electronic structure calculations indicate that the NaAlH{sub 4} and Na{sub 3}AlH{sub 6} structures are complex-ionic hydrides with Na{sup +} cations and AlH{sub 4}{sup -} and AlH{sub 6}{sup 3-} anions, respectively. Compound formation studies indicate the primary Ti-compound formed when doping the material at 33 at. % is TiAl{sub 3} , and likely Ti-Al compounds at lower doping rates. A general study of sorption kinetics of NaAlH{sub 4}, when doped with a variety of Ti-halide compounds, indicates a uniform response with the kinetics similar for all dopants. NMR multiple quantum studies of solution-doped samples indicate solvent interaction with the doped alanate. Raman spectroscopy was used to study the lattice dynamics of NaAlH{sub 4}, and illustrated the molecular ionic nature of the lattice as a separation of vibrational modes between the AlH{sub 4}{sup -} anion-modes and lattice-modes. In-situ Raman measurements indicate a stable AlH{sub 4}{sup -} anion that is stable at the melting temperature of NaAlH{sub 4}, indicating that Ti-dopants must affect the Al-H bond strength.

  17. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  18. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    SciTech Connect

    Barariu, G.; Giumanca, R.

    2006-07-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the

  19. Inverse sodium hydride: a theoretical study.

    PubMed

    Sawicka, Agnieszka; Skurski, Piotr; Simons, Jack

    2003-04-01

    A recent experimental investigation in which a salt containing the unusual charge distribution H(+) and Na(-) was synthesized and characterized prompted us to undertake an ab initio theoretical investigation. In the salt synthesized, the H(+) is bound to the nitrogen center of an amine and the Na(-) alkalide is "blocked" from approaching the protonated amine site by steric constraints of a cage structure. Although one expects that the Na(-) would deprotonate an unprotected R(3)N-H(+) cation, we decided to further explore this issue. Using extended atomic orbital basis sets and Møller-Plesset and coupled-cluster treatments of electron correlation, we examined the relative stabilities of the prototype (Me)(3)N + NaH, (Me)(3)N + Na(+) + H(-), (Me)(3)N-H(+) + Na(-), and (Me)(3)N-Na(+) + H(-) as well as the ion pair complexes (Me)(3)N-H(+).Na(-) and (Me)(3)N-Na(+).H(-). The primary focus of this effort was to determine whether the high-energy (Me)(3)N-H(+).Na(-) ion pair, which is the analogue of what the earlier workers termed "inverse sodium hydride", might be stable with respect to proton abstraction under any reasonable solvation conditions (which we treated within the polarized continuum model). Indeed, we find that such ion pairs are metastable (i.e., locally geometrically stable with a barrier to dissociation) for solvents having dielectric constants below approximately 2 but spontaneously decompose into their constituent ions for solvents with higher dielectric constants. We suggest that amines with large proton affinities and/or metals with weaker MH bond strengths should be explored experimentally. PMID:12656631

  20. REACTOR MODERATOR STRUCTURE

    DOEpatents

    Fraas, A.P.; Tudor, J.J.

    1963-08-01

    An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

  1. Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

    2013-03-01

    If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time

  2. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    SciTech Connect

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  3. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  4. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  5. Sodium-based hydrides for thermal energy applications

    NASA Astrophysics Data System (ADS)

    Sheppard, D. A.; Humphries, T. D.; Buckley, C. E.

    2016-04-01

    Concentrating solar-thermal power (CSP) with thermal energy storage (TES) represents an attractive alternative to conventional fossil fuels for base-load power generation. Sodium alanate (NaAlH4) is a well-known sodium-based complex metal hydride but, more recently, high-temperature sodium-based complex metal hydrides have been considered for TES. This review considers the current state of the art for NaH, NaMgH3- x F x , Na-based transition metal hydrides, NaBH4 and Na3AlH6 for TES and heat pumping applications. These metal hydrides have a number of advantages over other classes of heat storage materials such as high thermal energy storage capacity, low volume, relatively low cost and a wide range of operating temperatures (100 °C to more than 650 °C). Potential safety issues associated with the use of high-temperature sodium-based hydrides are also addressed.

  6. Sodium Reactor Experiment decommissioning. Final report

    SciTech Connect

    Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  7. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  8. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    NASA Astrophysics Data System (ADS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  9. Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs

    SciTech Connect

    Sienicki, J.J.; Horak, W.C.; Brookhaven National Lab., Upton, NY )

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.

  10. Kalkar nuclear power plant (SNR-300) - A sodium-cooled fast breeder reactor prototype

    SciTech Connect

    Morgenstern, F.H.

    1987-09-01

    The status of the Kalkar nuclear power plant in early summer 1986 is that, apart from later alterations to the workshop building, the assembly and non-nuclear commissioning work has practically been completed. From a technical point of view, nuclear commissioning of the plant can begin, but vital factors for this are the necessary nuclear licenses. The most important licensing prerequisites have been fulfilled;all essential appraisals have been available since January/February 1986. At the beginning of April 1986, the Reactor Safety Commission and the Radiation Protection Commission cast a positive vote for initial fuel loading. Before the accident in Chernobyl, but particularly since then, the issuing of the licenses has come under the political pressure of the commencing election campaign phase for the federal elections in January 1987. The initial project definition phase, the organizational boundary conditions, and the major requirements for the construction of the plant are summarized in chronological form. To provide the total picture, references dealing with general and technical aspects of the project are listed.

  11. Understanding the Origins of Nucleophilic Hydride Reactivity of a Sodium Hydride-Iodide Composite.

    PubMed

    Hong, Zonghan; Ong, Derek Yiren; Muduli, Subas Kumar; Too, Pei Chui; Chan, Guo Hao; Tnay, Ya Lin; Chiba, Shunsuke; Nishiyama, Yusuke; Hirao, Hajime; Soo, Han Sen

    2016-05-17

    Sodium hydride (NaH) has been commonly used as a Brønsted base in chemical syntheses, while it has rarely been employed to add hydride (H(-) ) to unsaturated electrophiles. We previously developed a procedure to activate NaH through the addition of a soluble iodide source and found that the new NaH-NaI composite can effect even stereoselective nucleophilic hydride reductions of nitriles, imines, and carbonyl compounds. In this work, we report that mixing NaH with NaI or LiI in tetrahydrofuran (THF) as a solvent provides a new inorganic composite, which consists of NaI interspersed with activated NaH, as revealed by powder X-ray diffraction, and both solid-state NMR and X-ray photoelectron spectroscopies. DFT calculations imply that this remarkably simple inorganic composite, which is comprised of NaH and NaI, gains nucleophilic hydridic character similar to covalent hydrides, resulting in unprecedented and unique hydride donor chemical reactivity. PMID:27038135

  12. FLUID MODERATED REACTOR

    DOEpatents

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  13. Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2009-07-01

    Analyses of supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be sure that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO{sub 2} Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 C core outlet temperature and a 470 C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact the

  14. Hydride Reduction by a Sodium Hydride–Iodide Composite

    PubMed Central

    Too, Pei Chui; Chan, Guo Hao; Tnay, Ya Lin

    2016-01-01

    Abstract Sodium hydride (NaH) is widely used as a Brønsted base in chemical synthesis and reacts with various Brønsted acids, whereas it rarely behaves as a reducing reagent through delivery of the hydride to polar π electrophiles. This study presents a series of reduction reactions of nitriles, amides, and imines as enabled by NaH in the presence of LiI or NaI. This remarkably simple protocol endows NaH with unprecedented and unique hydride‐donor chemical reactivity. PMID:26878823

  15. Incoherent Quasielastic Neutron Scattering study of hydrogen diffusion in thorium-zirconium hydrides

    NASA Astrophysics Data System (ADS)

    Terrani, Kurt A.; Mamontov, Eugene; Balooch, Mehdi; Olander, Donald R.

    2010-06-01

    Monophase thorium-zirconium hydrides (ThZr 2H x) have been fabricated starting from a metallic alloy and the hydrogen stoichiometry determined by X-ray diffraction. Incoherent Quasielastic Neutron Scattering (IQNS) on the hydrides was conducted over the temperature range 650-750 K at the Backscattering Silicon Spectrometer (BASIS) at the Spallation Neutron Source (SNS) at ORNL. The isotropic Chudley-Elliott model was utilized to analyze the quasielastic linewidth broadening data as function of momentum transfer. The diffusion coefficient and average jump distance of hydrogen atoms in ThZr 2H 5.6 and ThZr 2H 6.2 were extracted from the measurements.

  16. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  17. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

    1957-11-12

    This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

  18. Gas Diffusion in Metals: Fundamental Study of Helium-Point Defect Interactions in Iron and Kinetics of Hydrogen Desorption from Zirconium Hydride

    NASA Astrophysics Data System (ADS)

    Hu, Xunxiang

    The behavior of gaseous foreign species (e.g., helium and hydrogen), which are either generated, adsorbed or implanted within the structural materials (e.g., iron and zirconium) exposed to irradiation environments, is an important and largely unsolved topic, as they intensively interact with the irradiation-induced defects, or bond with the lattice atoms to form new compounds, and impose significant effects on their microstructural and mechanical properties in fission and fusion reactors. This research investigates two cases of gas diffusion in metals (i.e., the helium-point defect interactions in iron and kinetics of hydrogen desorption from zirconium hydride) through extensive experimental and modeling studies, with the objective of improving the understanding of helium effects on the microstructures of iron under irradiation and demonstrating the kinetics of hydrogen diffusion and precipitation behavior in zirconium that are crucial to predict cladding failures and hydride fuel performance. The study of helium effects in structural materials aims to develop a self-consistent, experimentally validated model of helium---point defect, defect cluster and intrinsic defects through detailed inter-comparisons between experimental measurements on helium ion implanted iron single crystals and computational models. The combination of thermal helium desorption spectrometry (THDS) experiment with the cluster dynamic model helps to reveal the influence of impurities on the energetics and kinetics of the He-defect interactions and to realize the identification of possible mechanisms governing helium desorption peaks. Positron annihilation spectroscopy is employed to acquire additional information on He-vacancy cluster evolution, which provides an opportunity to validate the model qualitatively. The inclusion of He---self-interstitial clusters extends the cluster dynamic model while MD simulations explore the effects of dislocation loops on helium clustering. In addition, the

  19. Zirconium hydride precipitation kinetics in Zircaloy-4 observed with synchrotron X-ray diffraction

    NASA Astrophysics Data System (ADS)

    Blackmur, M. S.; Robson, J. D.; Preuss, M.; Zanellato, O.; Cernik, R. J.; Shi, S.-Q.; Ribeiro, F.; Andrieux, J.

    2015-09-01

    High-energy synchrotron X-ray diffraction was used to investigate the isothermal precipitation of δ-hydride platelets in Zircaloy-4 at a range of temperatures relevant to reactor conditions, during both normal operation and thermal transients. From an examination of the rate kinetics of the precipitation process, precipitation slows with increasing temperature above 200 °C, due to a reduction in the thermodynamic driving force. A model for nucleation rate as a function of temperature was developed, to interpret the precipitation rates seen experimentally. While the strain energy associated with the misfit between hydrides and the matrix makes a significant contribution to the energy barrier for nucleation, a larger contribution arises from the interfacial energy. Diffusion distance calculations show that hydrogen is highly mobile in the considered thermal range and on the scale of inter-hydride spacing and it is not expected to be significantly rate limiting on the precipitation process that takes place under reactor operating conditions.

  20. Wall pressure exerted by hydrogenation of sodium aluminum hydride.

    SciTech Connect

    Perras, Yon E.; Dedrick, Daniel E.; Zimmerman, Mark D.

    2009-06-01

    Wall pressure exerted by the bulk expansion of a sodium aluminum hydride bed was measured as a function of hydrogen content. A custom apparatus was designed and loaded with sodium alanates at densities of 1.0, 1.1, and 1.16 g/cc. Four complete cycles were performed to identify variations in measured pressure. Results indicated poor correlation between exerted pressure and hydrogen capacity of the sodium alanate beds. Mechanical pressure due to the hydrogenation of sodium alanates does not influence full-scale system designs as it falls within common design factors of safety. Gas pressure gradients within the porous solid were identified and may limit reaction rates, especially for high aspect ratio beds.

  1. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate

  2. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOEpatents

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  3. Heterogeneous Transmutation Sodium Fast Reactor

    SciTech Connect

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  4. Ab initio-based fracture toughness estimates and transgranular traction-separation modelling of zirconium hydrides

    NASA Astrophysics Data System (ADS)

    Olsson, P. A. T.; Kese, K.; Kroon, M.; Alvarez Holston, A.-M.

    2015-06-01

    In this work we report the results of an ab initio study of the transgranular fracture toughness and cleavage of brittle zirconium hydrides. We use the Griffith-Irwin relation to assess the fracture toughness using calculated surface energy and estimated isotropic Voigt-Reuss-Hill averages of the elastic constants. The calculated fracture toughness values are found to concur well with experimental data, which implies that fracture is dominated by cleavage failure. To investigate the cleavage energetics, we model the decohesion process. To describe the interplanar interaction we adopt Rose’s universal binding energy relation, which is found to reproduce the behaviour accurately. The modelling shows that the work of fracture and ductility decreases with increasing hydrogen content.

  5. Complications from Dual Roles of Sodium Hydride as a Base and as a Reducing Agent

    PubMed Central

    Hesek, Dusan; Lee, Mijoon; Noll, Bruce C.; Fisher, Jed F.; Mobashery, Shahriar

    2012-01-01

    Sodium hydride is a common reagent for substrate activation in nucleophilic substitution reactions. Sodium hydride can behave both as a base and as a source of hydride. This dual ability in the presence of an electrophile such as benzyl bromide results in the formation of byproducts when dimethylformamide or acetonitrile are used as solvents for these reactions. The structural nature of these byproducts is revealed in this report. PMID:19215116

  6. The influence of stress state on the reorientation of hydrides in a zirconium alloy

    NASA Astrophysics Data System (ADS)

    Cinbiz, Mahmut N.; Koss, Donald A.; Motta, Arthur T.

    2016-08-01

    Hydride reorientation can occur in spent nuclear fuel cladding when subjected to a tensile hoop stress above a threshold value during cooling. Because in these circumstances the cladding is under a multiaxial stress state, the effect of stress biaxiality on the threshold stress for hydride reorientation is investigated using hydrided CWSR Zircaloy-4 sheet specimens containing ∼180 wt ppm of hydrogen and subjected to a two-cycle thermo-mechanical treatment. The study is based on especially designed specimens within which the stress biaxiality ratios range from uniaxial (σ2/σ1 = 0) to "near-equibiaxial" tension (σ2/σ1 = 0.8). The threshold stress is determined by mapping finite element calculations of the principal stresses and of the stress biaxiality ratio onto the hydride microstructure obtained after the thermo-mechanical treatment. The results show that the threshold stress (maximum principal stress) decreases from 155 to 75 MPa as the stress biaxiality increases from uniaxial to "near-equibiaxial" tension.

  7. Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels

    SciTech Connect

    Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki

    2007-07-01

    High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future SFR fuels. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and foreign fast reactors. The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to upgrade the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025. (authors)

  8. GAS COOLED NUCLEAR REACTORS

    DOEpatents

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  9. Multi-scale characterization of nanostructured sodium aluminum hydride

    NASA Astrophysics Data System (ADS)

    NaraseGowda, Shathabish

    Complex metal hydrides are the most promising candidate materials for onboard hydrogen storage. The practicality of this class of materials is counter-poised on three critical attributes: reversible hydrogen storage capacity, high hydrogen uptake/release kinetics, and favorable hydrogen uptake/release thermodynamics. While a majority of modern metallic hydrides that are being considered are those that meet the criteria of high theoretical storage capacity, the challenges lie in addressing poor kinetics, thermodynamics, and reversibility. One emerging strategy to resolve these issues is via nanostructuring or nano-confinement of complex hydrides. By down-sizing and scaffolding them to retain their nano-dimensions, these materials are expected to improve in performance and reversibility. This area of research has garnered immense interest lately and there is active research being pursued to address various aspects of nanostructured complex hydrides. The research effort documented here is focused on a detailed investigation of the effects of nano-confinement on aspects such as the long range atomic hydrogen diffusivities, localized hydrogen dynamics, microstructure, and dehydrogenation mechanism of sodium alanate. A wide variety of microporous and mesoporous materials (metal organic frameworks, porous silica and alumina) were investigated as scaffolds and the synthesis routes to achieve maximum pore-loading are discussed. Wet solution infiltration technique was adopted using tetrahydrofuran as the medium and the precursor concentrations were found to have a major role in achieving maximum pore loading. These concentrations were optimized for each scaffold with varying pore sizes and confinement was quantitatively characterized by measuring the loss in specific surface area. This work is also aimed at utilizing neutron and synchrotron x-ray characterization techniques to study and correlate multi-scale material properties and phenomena. Some of the most advanced

  10. Development of Continuous Cooling Transformation Diagrams of Zirconium-Niobium Alloy Phase Transformations within the Kolmogorov-Johnson-Mehl-Avrami Framework

    NASA Astrophysics Data System (ADS)

    Kautz, Elizabeth J.

    Microstructure and chemistry of zirconium alloys have a major influence on material performance, including mechanical properties and corrosion resistance. Therefore, it is essential to thoroughly understand processing required to obtain desired microstructures for application in commercial nuclear reactors. Zirconium-niobium alloys are of particular interest for commercial nuclear applications (e.g., in boiling water reactors, pressurized water reactors, Canadian deuterium uranium reactors) due to increased corrosion resistance in aqueous environments over other zirconium alloys. Heat treatments of zirconium-niobium alloys affect overall microstructure, precipitate distributions and size, and ultimately determine material performance. Phase transformations in zirconium-niobium alloys were modeled for a range of niobium concentrations and heat treatment conditions, by conducting controlled experiments. Heat-flux differential scanning calorimetry was performed and data was collected and analyzed for zirconium-niobium alloys with niobium content ranging from 0.6-3.0 weight percent. Continuous cooling transformation diagrams were constructed for slow cooling rate conditions (9-34°C/minute) based on calorimetry test results. A standard operating procedure for performing these calorimetry tests and corresponding data analysis technique was developed specifically to study the zirconium-niobium material system. A mathematical model was developed utilizing the Kolmogorov-Johnson-Mehl-Avrami theory that accurately describes phase transformations upon continuous cooling in zirconium-niobium binary alloys. This model relates fraction of phase transformed to kinetic parameters that were calculated from experimental test results in order to model the phase transformation for various cooling rates from 10-40°C/minute.

  11. VELM61 and VELM22: Multigroup cross-section libraries for sodium-cooled reactor shield analysis

    SciTech Connect

    Fu, C.Y.; Ingersoll, D.T.

    1987-04-01

    Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23..gamma.. and 22n/10..gamma.., are subsets of the Vitamin-E 174n/38..gamma.. group structure, and are tailored to the iron and sodium resonances, windows, and capture gamma-ray spectra. Each of the two libraries are available in two formats, the AMPX master format and the ANISN format. Cross sections for all materials in the Vitamin-E library were collapsed using a standard energy weighting function, and in addition, several cross-section sets for each of the major constituents of commercial grade sodium, stainless steel (types 304 and 316), and carbon steel were derived using several problem-dependent weighting functions for averaging the fine groups. Effects of various group structures and weighting functions on the accuracy of the broad group libraries are studied by ANISN analysis of a typical sodium-iron shield configuration.

  12. Retrospective Reactor Dosimetry with Zirconium Alloy Samples

    SciTech Connect

    Greenwood, Lawrence R.; Foster, John P.

    2009-11-01

    Retrospective measurements are routinely performed with stainless steel samples. Recent experiments have been successfully conducted using zirconium alloy samples, involving somewhat different neuron activation reactions than are normally encountered with stainless steel samples. The alloy composition consisted of nominally 1% (by weight) niobium, 1% tin, and 0.1% iron, with the balance zirconium. The activation products observed in the samples by gamma spectroscopy included Zr-95, Nb-95, Sn-113, Sb-125, Mn-54, Co-60, Nb-94, and Ta-182. The niobium was then chemically separated following ASTM procedure E1297 and the Nb-93m activities were measured by x-ray spectroscopy. The thermal neutron fluences, as determined independently by the neutron capture gamma reactions to Zr-95, Sn-113, Nb-94, and Sn/Sb-125, were in excellent agreement. The fast neutron fluences, as determined separately by the Fe-54(n,p)Mn-54 and Nb-93(n,n’)Nb-93m reactions, were also in good agreement, thus demonstrating the versatility of the retrospective dosimetry technique.

  13. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  14. (Gas-cooled reactor materials)

    SciTech Connect

    Rittenhouse, P.L.

    1988-06-30

    The meeting of the managers of the US/FRG/CH cooperative subprogram on materials for gas-cooled reactors is described and the status of each of the work packages comprising this cooperation is summarized. Four proposals for new areas of cooperative work were developed. Briefings by sponsoring organizations on the status of gas-cooled reactor programs in the FRG are discussed and experimental efforts being conducted at KFA on materials are reviewed.

  15. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  16. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  17. Small Liquid Metal Cooled Reactor Safety Study

    SciTech Connect

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  18. Thermal hydraulic analysis of two-phase closed thermosyphon cooling system for new cold neutron source moderator of Breazeale research reactor at Penn State

    NASA Astrophysics Data System (ADS)

    Habte, Melaku

    A cold neutron source cooling system is required for the Penn State's next generation cold neutron source facility that can accommodate a variable heat load up to about ˜10W with operating temperature of about 28K. An existing cold neutron source cooling system operating at the University of Texas Cold Neutron Source (TCNS) facility failed to accommodate heat loads upwards of 4W with the moderator temperature reaching a maximum of 44K, which is the critical temperature for the operating fluid neon. The cooling system that was used in the TCNS cooling system was a two-phase closed thermosyphon with a reservoir (TPCTR). The reservoir containing neon gas is kept at room temperature. In this study a detailed thermal analysis of the fundamental operating principles of a TPCTR were carried out. A detailed parametric study of the various geometric and thermo-physical factors that affect the limits of the operational capacity of the TPCTR investigated. A CFD analysis is carried out in order to further refine the heat transfer analysis and understand the flow structure inside the thermosyphon and the two-phase nucleate boiling in the evaporator section of the thermosyphon. In order to help the new design, a variety of ways of increasing the operating range and heat removal capacity of the TPCTR cooling system were analyzed so that it can accommodate the anticipated heat load of 10W or more. It is found, for example, that doubling the pressure of the system will increase the capacity index zeta by 50% for a system with an initial fill ratio FR of 1. A decrease in cryorefrigeration performance angle increases the capacity index. For example taking the current condition of the TCNS system and reducing the angle from the current value of ˜700 by half (˜350) will increase the cooling power 300%. Finally based on detailed analytic and CFD analysis the best operating condition were proposed.

  19. Computational study of sodium magnesium hydride for hydrogen storage applications

    NASA Astrophysics Data System (ADS)

    Soto Valle, Fernando Antonio

    Hydrogen offers considerable potential benefits as an energy carrier. However, safe and convenient storage of hydrogen is one of the biggest challenges to be resolved in the near future. Sodium magnesium hydride (NaMgH 3) has attracted attention as a hydrogen storage material due to its light weight and high volumetric hydrogen density of 88 kg/m3. Despite the advantages, hydrogen release in this material occurs at approximately 670 K, which is well above the operable range for on-board hydrogen storage applications. In this regard, hydrogen release may be facilitated by substitution doping of transition-metals. This dissertation describes first-principles computational methods that enable an examination of the hydrogen storage properties of NaMgH3. The novel contribution of this dissertation includes a combination of crystal, supercell, and surface slab calculations that provides new and relevant insights about the thermodynamic and kinetic properties of NaMgH3. First-principles calculations on the pristine crystal structure provide a starting reference point for the study of this material as a hydrogen storage material. To the best of our knowledge, it is reported for the first time that a 25% mol doping concentration of Ti, V, Cu, and Zn dopants reduce the reaction enthalpy of hydrogen release for NaMgH3. The largest decrease in the DeltaH(298 K) value corresponds to the Zn-doped model (67.97 kJ/(mol H2)). Based on cohesive energy calculations, it is reported that at the 6.25% mol doping concentration, Ti and Zn dopants are the only transition metals that destabilize the NaMgH3 hydride. In terms of hydrogen removal energy, it is quantified that the energy cost to remove a single H from the Ti-doped supercell model is 0.76 eV, which is lower with respect to the pristine model and other prototypical hydrogen storage materials. From the calculation of electronic properties such as density of states, electron density difference, and charge population analysis

  20. Aqueous sodium borohydride induced thermally stable porous zirconium oxide for quick removal of lead ions

    PubMed Central

    Nayak, Nadiya B.; Nayak, Bibhuti B.

    2016-01-01

    Aqueous sodium borohydride (NaBH4) is well known for its reducing property and well-established for the development of metal nanoparticles through reduction method. In contrary, this research paper discloses the importance of aqueous NaBH4 as a precipitating agent towards development of porous zirconium oxide. The boron species present in aqueous NaBH4 play an active role during gelation as well as phase separated out in the form of boron complex during precipitation, which helps to form boron free zirconium hydroxide [Zr(OH)4] in the as-synthesized condition. Evolved in-situ hydrogen (H2) gas-bubbles also play an important role to develop as-synthesized loose zirconium hydroxide and the presence of intra-particle voids in the loose zirconium hydroxide help to develop porous zirconium oxide during calcination process. Without any surface modification, this porous zirconium oxide quickly adsorbs almost hundred percentages of toxic lead ions from water solution within 15 minutes at normal pH condition. Adsorption kinetic models suggest that the adsorption process was surface reaction controlled chemisorption. Quick adsorption was governed by surface diffusion process and the adsorption kinetic was limited by pore diffusion. Five cycles of adsorption-desorption result suggests that the porous zirconium oxide can be reused efficiently for removal of Pb (II) ions from aqueous solution. PMID:26980545

  1. Aqueous sodium borohydride induced thermally stable porous zirconium oxide for quick removal of lead ions.

    PubMed

    Nayak, Nadiya B; Nayak, Bibhuti B

    2016-01-01

    Aqueous sodium borohydride (NaBH4) is well known for its reducing property and well-established for the development of metal nanoparticles through reduction method. In contrary, this research paper discloses the importance of aqueous NaBH4 as a precipitating agent towards development of porous zirconium oxide. The boron species present in aqueous NaBH4 play an active role during gelation as well as phase separated out in the form of boron complex during precipitation, which helps to form boron free zirconium hydroxide [Zr(OH)4] in the as-synthesized condition. Evolved in-situ hydrogen (H2) gas-bubbles also play an important role to develop as-synthesized loose zirconium hydroxide and the presence of intra-particle voids in the loose zirconium hydroxide help to develop porous zirconium oxide during calcination process. Without any surface modification, this porous zirconium oxide quickly adsorbs almost hundred percentages of toxic lead ions from water solution within 15 minutes at normal pH condition. Adsorption kinetic models suggest that the adsorption process was surface reaction controlled chemisorption. Quick adsorption was governed by surface diffusion process and the adsorption kinetic was limited by pore diffusion. Five cycles of adsorption-desorption result suggests that the porous zirconium oxide can be reused efficiently for removal of Pb (II) ions from aqueous solution. PMID:26980545

  2. Aqueous sodium borohydride induced thermally stable porous zirconium oxide for quick removal of lead ions

    NASA Astrophysics Data System (ADS)

    Nayak, Nadiya B.; Nayak, Bibhuti B.

    2016-03-01

    Aqueous sodium borohydride (NaBH4) is well known for its reducing property and well-established for the development of metal nanoparticles through reduction method. In contrary, this research paper discloses the importance of aqueous NaBH4 as a precipitating agent towards development of porous zirconium oxide. The boron species present in aqueous NaBH4 play an active role during gelation as well as phase separated out in the form of boron complex during precipitation, which helps to form boron free zirconium hydroxide [Zr(OH)4] in the as-synthesized condition. Evolved in-situ hydrogen (H2) gas-bubbles also play an important role to develop as-synthesized loose zirconium hydroxide and the presence of intra-particle voids in the loose zirconium hydroxide help to develop porous zirconium oxide during calcination process. Without any surface modification, this porous zirconium oxide quickly adsorbs almost hundred percentages of toxic lead ions from water solution within 15 minutes at normal pH condition. Adsorption kinetic models suggest that the adsorption process was surface reaction controlled chemisorption. Quick adsorption was governed by surface diffusion process and the adsorption kinetic was limited by pore diffusion. Five cycles of adsorption-desorption result suggests that the porous zirconium oxide can be reused efficiently for removal of Pb (II) ions from aqueous solution.

  3. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  4. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E.

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  5. Temperature dependences of the delayed hydride cracking rate of fuel claddings made of zirconium alloys of various compositions

    NASA Astrophysics Data System (ADS)

    Markelov, V. A.; Gusev, A. Yu.; Kotov, P. V.; Novikov, V. V.; Saburov, N. S.

    2014-04-01

    The temperature dependences of the delayed hydride cracking (DHC) rate of Zr-1Nb and Zr-0.8Nb-0.8Sn-0.3Fe alloy claddings are studied in the range 127-300°C in comparison with the data obtained for Zr-2.5Nb and Zircaloy-4 alloys earlier. The samples are in the state of cold deformation and stress relief at 400°C for 24 h and in the state of preliminary hydrogen saturation to a hydrogen concentration of 0.02 wt %. As the strength of a zirconium alloy decreases and its ductility increases, the DHC rate and its high-temperature limit for a linear Arrhenius equation decreases, and the fractographic patterns of the fracture surfaces are different.

  6. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  7. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  8. REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING WATER PIPES, COOLING AIR DUCTS, AND SHIELDING. INL NEGATIVE NO. 776. Unknown Photographer, 10/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  9. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  10. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1995-01-01

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  11. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  12. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  13. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

    SciTech Connect

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

    2014-11-18

    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  14. MEANS FOR SHIELDING AND COOLING REACTORS

    DOEpatents

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  15. Gas-cooled nuclear reactor

    DOEpatents

    Peinado, Charles O.; Koutz, Stanley L.

    1985-01-01

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

  16. Synthesis of zirconium carbide from zircon concentrates in a thermal plasma reactor

    SciTech Connect

    Vidal, E.E.; Taylor, P.R.; Manrique, M.

    1996-10-01

    Ultrafine zirconium carbide powder has been synthesized using zircon concentrate and methane in a thermal plasma reactor. The effect of several experimental variables, such as power used and zircon particle size, on the conversion has been studied. A thermodynamic analysis of the possible reactions and stable compounds was performed. A mathematical model that describes the in-flight decomposition of zircon particles was developed based on a description of the temperature and velocity profiles calculated in the reactor. It was shown that the formation of zirconium carbide from zircon and methane is thermodynamically possible. Complete vaporization of zircon in the reactor is predicted when the particle size is below 45 {micro}m and the effective power entering the reactor is 15 kW. The powders were collected and characterized by XRD, SEM, TEM and chemical analysis using ICP. Results of the product analysis indicated the presence of zirconium carbide, zirconium oxide, silicon carbide and silicon oxide.

  17. Safety Characteristics of LBE Cooled Long-Life Small Reactor, 'LSPR'

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino

    2002-07-01

    Lead bismuth eutectic (LBE) shows a good performance on neutron economy, and LBE cooled fast reactor can be designed as an excellent long-life small reactor. LBE is good not only for neutron economy but for chemical inertness and high boiling point, which may realize a much safer reactor than conventional sodium-cooled reactor. We have designed such a long-life small reactor and name it LSPR. This paper presents safety characteristics of LSPR. (authors)

  18. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  19. Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors

    SciTech Connect

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-07-01

    This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

  20. Zirconium

    USGS Publications Warehouse

    Bedinger, G.M.

    2013-01-01

    Zirconium is the 20th most abundant element in the Earth’s crust. It occurs in a variety of rock types and geologic environments but most often in igneous rocks in the form of zircon (ZrSiO4). Zircon is recovered as a coproduct of the mining and processing of heavy mineral sands for the titanium minerals ilmenite and rutile. The sands are formed by the weathering and erosion of rock containing zircon and titanium heavy minerals and their subsequent concentration in sedimentary systems, particularly in coastal environments. A small quantity of zirconium, less than 10 kt/a (11,000 stpy), compared with total world production of 1.4 Mt (1.5 million st) in 2012, was derived from the mineral baddeleyite (ZrO2), produced from a single source in Kovdor, Russia.

  1. Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining

    SciTech Connect

    J. A. Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman

    2005-09-01

    The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.

  2. Terminal cool-down temperature-dependent hydride reorientations in Zr-Nb Alloy claddings under dry storage conditions

    NASA Astrophysics Data System (ADS)

    Min, Su-Jeong; Won, Ju-Jin; Kim, Kyu-Tae

    2014-05-01

    In order to simulate high-burnup fuel cladding degradation under various interim dry storage conditions, 250 ppm and 500 ppm hydrogen-charged Zr-Nb alloy cladding tubes were used to investigate the effect of terminal cool-down temperature on hydride reorientations and subsequent mechanical property degradations under a tensile hoop stress of 150 MPa with two cooling rates of 2.0 and 7.0 °C/min from a peak temperature of 400 °C to three respective terminal cool-down temperatures of 300, 200 and 25 °C. The cool-down tests showed that the slower cooling rate, the lower terminal cool-down temperature and the higher hydrogen content generated the larger fraction of radial hydrides precipitated during the cool-down. This may be explained by hydrogen solid solubilities for precipitation at the respective terminal cool-down temperatures, by cooling rate-dependent residence times at a relatively high temperature during the cool-down and by remaining circumferential hydrides prior to the cool-down. Ultimate tensile strengths, plastic strains and fracture modes for the tensile-tested specimens are found to be well correlated to the amount of the radial and circumferential hydrides and hydride morphologies.

  3. Cooling rate- and hydrogen content-dependent hydride reorientation and mechanical property degradation of Zr-Nb alloy claddings

    NASA Astrophysics Data System (ADS)

    Min, Su-Jeong; Kim, Myeong-Su; Kim, Kyu-Tae

    2013-10-01

    As-received, 250 ppm and 500 ppm hydrogen-charged Zr-Nb alloy cladding tubes were employed to evaluate the effects of cooling rates on the extent of radial hydride formation and mechanical property degradations using the cooling rates of 0.3, 2.0, 4.0 and 15.0 °C/min from 400 °C to room temperature under a tensile hoop stress of 150 MPa, which may simulate a long-term cladding cool-down during an interim dry storage. The hydrogen-charged claddings showed the larger fraction and the longer average length of the radial hydrides with the slower cooling rate. These effects are more dominant for the 250 ppm-H cladding tubes than for the 500 ppm-H cladding tubes. The cooling rate- and the hydrogen content-dependent radial hydride morphology may be explained by the difference in the remaining circumferential hydride fraction at 400 °C and the cooling rate-dependent hydride nucleation and growth rates. The extent of the mechanical property degradations for the hydrogen-charged cladding tubes becomes larger with the decrease of the cooling rate and for the 250 ppm-H specimens. These phenomena are well correlated with the cooling rate- and the hydrogen content-dependent radial hydride fraction and length.

  4. Helium-cooled high temperature reactors

    SciTech Connect

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  5. Optimally moderated nuclear fission reactor and fuel source therefor

    DOEpatents

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  6. Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

    NASA Astrophysics Data System (ADS)

    Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

    Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

  7. Cooling system for a nuclear reactor

    DOEpatents

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  8. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  9. Electrochemical research in chemical hydrogen storage materials: Sodium borohydride and organotin hydrides

    NASA Astrophysics Data System (ADS)

    McLafferty, Jason

    Chemical storage of hydrogen involves release of hydrogen in a controlled manner from materials in which the hydrogen is covalently bound. Sodium borohydride and aminoborane are two materials given consideration as chemical hydrogen storage materials by the US Department of Energy. A very significant barrier to adoption of these materials as hydrogen carriers is their regeneration from "spent fuel," i.e., the material remaining after discharge of hydrogen. In this thesis, some research directed at regeneration of sodium borohydride and aminoborane is described. For sodium borohydride, electrochemical reduction of boric acid and sodium metaborate (representing spent fuel) in alkaline, aqueous solution has been investigated. Similarly to literature reports (primarily patents), a variety of cathode materials were tried in these experiments. Additionally, approaches directed at overcoming electrostatic repulsion of borate anion from the cathode, not described in the previous literature for electrochemical reduction of spent fuels, have been attempted. A quantitative analytical method for measuring the concentration of sodium borohydride in alkaline aqueous solution has been developed as part of this work and is described herein. Finally, findings from stability tests for sodium borohydride in aqueous solutions of several different compositions are reported. For aminoborane, other research institutes have developed regeneration schemes involving tributyltin hydride. In this thesis, electrochemical reduction experiments attempting to regenerate tributyltin hydride from tributyltin chloride (a representative by-product of the regeneration scheme) are described. These experiments were performed in the non-aqueous solvents acetonitrile and 1,2-dimethoxyethane. A non-aqueous reference electrode for electrolysis experiments in acetonitrile was developed and is described.

  10. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  11. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  12. Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

    NASA Astrophysics Data System (ADS)

    Hsu, Hsiao-Hung; Tsay, Leu-Wen

    2011-01-01

    Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.

  13. PROCESS FOR COOLING A NUCLEAR REACTOR

    DOEpatents

    Borst, L.B.

    1962-12-11

    This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

  14. Status and perspective of development of cold moderators at the IBR-2 reactor

    NASA Astrophysics Data System (ADS)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  15. Transmission Electron Microscopy Studies on Titanium-doped Sodium Aluminum Hydride

    NASA Astrophysics Data System (ADS)

    Culnane, Lance F.

    Hydrogen fuel cells play an important role in today's diverse and blossoming alternative energy industry. One of the greatest technological barriers for vehicular applications is the storage of hydrogen (which is required to power hydrogen fuel cells). Storing hydrogen as a gas is not volume efficient, and storing it as a liquid is not cost effective, therefore solid-state storage of hydrogen, such as in metal hydrides offers the most potential for success since many metal hydrides have attractive qualities for hydrogen storage such as: high volumetric capacity, cost efficiency, weight efficiency, low refueling times, and most importantly, high safety. Unfortunately, a compound has not been discovered which contains all of the attractive hydrogen storage qualities for vehicular applications. Sodium aluminum hydride (NaAlH 4) is one of the few compounds which is close to meeting requirements for car manufacturers, and has perhaps been researched the most extensively out of all metal hydrides in the last 15 years. This arises from the remarkable discovery by Bogdanovic who found that doping NaAlH4 with Ti dopants enabled the reversible dehydrogenation and hydrogenation of NaAlH 4 at mild conditions. Various evidence and theories have been proposed to suggest explanations for the enhanced kinetic effect that Ti-doping and ball-milling provide. However, the research community has not reached a consensus as to the exact role of Ti-dopants. If the role of titanium in the NaAlH4 dehydrogenation/hydrogenation mechanism could be understood, then more attractive metal hydrides could be designed. To this end, we conducted Transmission Electron Microscopy (TEM) studies to explain the role of the Ti dopants. The first known thorough particle size analysis of the NaAlH4 system was conducted, as well as TEM-EELS (Electron Energy Loss Spectroscopy), TEM-EDS (Energy Dispersive X-ray Spectroscopy), and in-situ imaging studies. Preparation methods were found to be important for the

  16. Space-reactor system and subsystem investigations: Cost and schedule estimates for reactor and shield subsystems technology development. SP-100 program

    NASA Astrophysics Data System (ADS)

    Determan, W. R.; Harty, R. B.; Hylin, C.

    1983-06-01

    Cost and schedule estimates of the technology development for reactor and shielding subsystems of a 100-kWe class space reactor electric system are presented. The subsystems technology development (which includes reactor and shield subsystems ground testing) is supported by materials and processes development and component development. For the purpose of the cost estimate, seven generic types of reactor subsystems were used: uranium-zirconium hydride, NaK-cooled thermal reactor; lithium-cooled, refractory-clad fast reactors; Na- or K-cooled fast reactor; in-core thermionic reactor; inert gas-cooled particle fuel reactor; inert gas-cooled metal-clad fast reactor; and heat pipe-cooled fast reactor. Also three levels of technology were included for each of the generic types of reactor subsystem: current, improved, and advanced. The data in this report encompass all these technology levels. The shielding subsystem uses both gamma (heavy-metal) and neutron (hydrogenous material) shields. The shields considered would be used in conjunction with unmanned payloads.

  17. Metal Hydride and Alkali Halide Opacities in Extrasolar Giant Planets and Cool Stellar Atmospheres

    NASA Technical Reports Server (NTRS)

    Weck, Philippe F.; Stancil, Phillip C.; Kirby, Kate; Schweitzer, Andreas; Hauschildt, Peter H.

    2006-01-01

    The lack of accurate and complete molecular line and continuum opacity data has been a serious limitation to developing atmospheric models of cool stars and Extrasolar Giant Planets (EGPs). We report our recent calculations of molecular opacities resulting from the presence of metal hydrides and alkali halides. The resulting data have been included in the PHOENIX stellar atmosphere code (Hauschildt & Baron 1999). The new models, calculated using spherical geometry for all gravities considered, also incorporate our latest database of nearly 670 million molecular lines, and updated equations of state.

  18. In-Bed Accountability Development for a Passively Cooled, Electrically Heated Hydride (PACE) Bed

    SciTech Connect

    Klein, J.E.

    2005-07-15

    A nominal 1500 STP-L PAssively Cooled, Electrically heated hydride (PACE) Bed has been developed for implementation into a new Savannah River Site tritium project. The 1.2 meter (four-foot) long process vessel contains on internal 'U-tube' for tritium In-Bed Accountability (IBA) measurements. IBA will be performed on six, 12.6 kg production metal hydride storage beds.IBA tests were done on a prototype bed using electric heaters to simulate the radiolytic decay of tritium. Tests had gas flows from 10 to 100 SLPM through the U-tube or 100 SLPM through the bed's vacuum jacket. IBA inventory measurement errors at the 95% confidence level were calculated using the correlation of IBA gas temperature rise, or (hydride) bed temperature rise above ambient temperature, versus simulated tritium inventory.Prototype bed IBA inventory errors at 100 SLPM were the largest for gas flows through the vacuum jacket: 15.2 grams for the bed temperature rise and 11.5 grams for the gas temperature rise. For a 100 SLPM U-tube flow, the inventory error was 2.5 grams using bed temperature rise and 1.6 grams using gas temperature rise. For 50 to 100 SLPM U-tube flows, the IBA gas temperature rise inventory errors were nominally one to two grams that increased above four grams for flows less than 50 SLPM. For 50 to 100 SLPM U-tube flows, the IBA bed temperature rise inventory errors were greater than the gas temperature rise errors, but similar errors were found for both methods at gas flows of 20, 30, and 40 SLPM.Electric heater IBA tests were done for six production hydride beds using a 45 SLPM U-tube gas flow. Of the duplicate runs performed on these beds, five of the six beds produced IBA inventory errors of approximately three grams: consistent with results obtained in the laboratory prototype tests.

  19. In-Bed Accountability Development for a Passively Cooled, Electrically Heated Hydride (PACE) Bed

    SciTech Connect

    KLEIN, JAMES

    2004-10-12

    A nominal 1500 STP-L PAssively Cooled, Electrically heated hydride (PACE) Bed has been developed for implementation into a new Savannah River Site tritium project. The 1.2 meter (four-foot) long process vessel contains an internal ''U-tube'' for tritium In-Bed Accountability (IBA) measurements. IBA will be performed on six, 12.6 kg production metal hydride storage beds. IBA tests were done on a prototype bed using electric heaters to simulate the radiolytic decay of tritium. Tests had gas flows from 10 to 100 SLPM through the U-tube or 100 SLPM through the bed's vacuum jacket. IBA inventory measurement errors at the 95 percent confidence level were calculated using the correlation of IBA gas temperature rise, or (hydride) bed temperature rise above ambient temperature, versus simulated tritium inventory. Prototype bed IBA inventory errors at 100 SLPM were the largest for gas flows through the vacuum jacket: 15.2 grams for the bed temperature rise and 11.5 grams for the gas temperature rise. For a 100 SLPM U-tube flow, the inventory error was 2.5 grams using bed temperature rise and 1.6 grams using gas temperature rise. For 50 to 100 SLPM U-tube flows, the IBA gas temperature rise inventory errors were nominally one to two grams that increased above four grams for flows less than 50 SLPM. For 50 to 100 SLPM U-tube flows, the IBA bed temperature rise inventory errors were greater than the gas temperature rise errors, but similar errors were found for both methods at gas flows of 20, 30, and 40 SLPM. Electric heater IBA tests were done for six production hydride beds using a 45 SLPM U-tube gas flow. Of the duplicate runs performed on these beds, five of the six beds produced IBA inventory errors of approximately three grams: consistent with results obtained in the laboratory prototype tests.

  20. Passive Reactor Cooling Using Capillary Porous Wick

    SciTech Connect

    Miller, Christopher G.; Lin, Thomas F.

    2006-07-01

    Long-term reliability of actively pumped cooling systems is a concern in space-based nuclear reactors. Capillary-driven passive cooling systems are being considered as an alternative to gravity-driven systems. The high surface tension of liquid lithium makes it attractive as the coolant in a capillary-driven cooling system. A system has been conceived in which the fuel rod of a reactor is surrounded by a concentric wick through which liquid lithium flows to provide cooling under normal and emergency operating conditions. Unheated wicking experiments at three pressures using four layered screen mesh wicks of different porosities and three relatively high surface tension fluids have been conducted to gain insight into capillary phenomena for such a capillary cooling system. All fluids tested demonstrated wicking ability in each of the wick structures for all pressures, and wicking ability for each fluid increased with decreasing wick pore size. An externally heated wicking experiment with liquid lithium as the wicking fluid was also conducted. In addition to wicking experiments, a heater rod is under development to simulate the fuel rod of a space based nuclear reactor by providing a heat flux of up to 110 kW/m{sup 2}. Testing of this heater rod has shown its ability to undergo repeated cycling from below 533 K to over 1255 K without failure. This heater rod will be integrated into lithium wicking experiments to provide more realistic simulation of the proposed capillary-driven space nuclear reactor cooling system. (authors)

  1. The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

    NASA Astrophysics Data System (ADS)

    Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

    2013-10-01

    A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

  2. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system.

  3. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    SciTech Connect

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  4. Measured and calculated fast neutron spectra in a depleted uranium and lithium hydride shielded reactor

    NASA Technical Reports Server (NTRS)

    Lahti, G. P.; Mueller, R. A.

    1973-01-01

    Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.

  5. Processes influencing cooling of reactor effluents

    SciTech Connect

    Magoulas, V.E.; Murphy, C.E. Jr.

    1982-06-07

    Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures.

  6. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    SciTech Connect

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

    2006-12-12

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

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

    DOEpatents

    Mariani, Robert Dominick

    2014-09-09

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

  8. Defining the stages of annealing in a moderately deformed commercial Zirconium alloy

    NASA Astrophysics Data System (ADS)

    Kumar, Gulshan; Singh, Ramesh; Singh, Jaiveer; Srivastava, D.; Dey, G. K.; Samajdar, I.

    2015-11-01

    Fully recrystallized Zircaloy-4 was cold rolled to 20% reduction in thickness. The deformed microstructure had fragmented and non-fragmented grains. Fragmentation represented deformation-induced refinement in grain size. Typically, the fragmented grains had more misorientation and were finer than the as-received grains. The deformed samples were subjected to 650°C annealing for different time periods, followed by water quenching. Based on experimental observations, three distinct stages of annealing were noted. Stage I caused changes in the misorientations of the non-fragmented grains, while the fragmented regions remained unaffected. This was also the most effective stage for residual stress relief. In stage II, discontinuous recrystallization and grain coarsening consumed the fragmented regions. This stage provided the highest softening. Finally, stage III created recovery-induced grain refinement of the larger non-fragmented grains. A combination of indirect and direct observations thus provided a complete picture of the annealing related microstructural changes in a moderately deformed commercial Zirconium alloy.

  9. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  10. Liquid-metal-cooled reactor

    DOEpatents

    Hutter, E.

    A perforated depressor plate extending across the bottom of the instrument tree of a fast breeder reactor cooperates with a circular cylindrical metal bellows forming a part of the upper adapter of each core assembly and bearing on the bottom of the depressor plate to restrict flow of coolant between core assemblies, thereby reducing significantly the pressure differential between the coolant inside the core assemblies and the coolant outside of the core assemblies. Openings in the depressor plate are slightly smaller than the top of the upper adapter so the depressor plate will serve as a backup mechanical holddown for the core. In addition, coolant mixing devices and locating devices are provided attached to the depressor plate.

  11. The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

    SciTech Connect

    Mann, A.; Herrick, R.; Gunn, J.; Husband, W.; Smith, M.; Fletcher, B.

    2007-07-01

    Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal of the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)

  12. Risk Management for Sodium Fast Reactors.

    SciTech Connect

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  13. Comparison of codes and neutron IC data used in US and Russia for the Topaz-II nuclear reactor assessment

    SciTech Connect

    Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Gomin, Y.A.; Maiorov, L.V.; Lobynstev, V.A.; Polyakov, D.N.; Sapir, J.; Streetman, J.R.

    1993-11-01

    Topaz-II is a heterogeneous, epithermal reactor, fueled with highly enriched uranium-dioxide, cooled with NaK, and moderated with zirconium-hydride. The reactor core contains 37 single-cell thermionic fuel elements, and is surrounded by a radial beryllium reflector that contains 12 rotatable control drums with poison segments. For the physics analysis of TOPAZ II it is necessary to use the Monte Carlo method. The United States (US) and Russia used two different Monte Carlo codes, namely MCNP and MCU-2, respectively. The work described in this paper was aimed at comparing the codes and neutronic data used in the US and Russia for verification of Topaz-II nuclear safety. For this purpose, the US and Russia developed a joint benchmark model of the Topaz-II reactor. The American and Russian teams performed independent computations for a series of variants representing potential water immersion accidents. Comparison of the MCNP and MCU-2 codes showed somewhat different results both for the absolute values of k{sub eff} and for reactivity effects. Future calculations will be performed to obtain a detailed understanding of the reasons for such discrepancies. For these analyses it will be necessary for the US and Russian teams to exchange neutronic data on Topaz-II physics calculations.

  14. Hydrogen generation from hydrides in millimeter scale reactors for micro proton exchange membrane fuel cell applications

    NASA Astrophysics Data System (ADS)

    Zhu, L.; Kim, D.; Kim, H.; Masel, R. I.; Shannon, M. A.

    This paper introduces and discusses the feasibility of millimeter scale powder packed-bed reactors using high energy density chemical hydrides for micro-PEM fuel cell applications. Two different reactors were designed and tested using LiBH 4, LiAlH 4, NaAlH 4 and CaH 2 hydride fuels. The mechanisms that limit the total yield of H 2 generated and impact the performance of the hydrogen generator are investigated in this paper, including density, solubility and porosity of the reaction byproducts. The volume expansion of the byproducts has significant effect on the total energy density of micro-hydrogen generators because the byproducts are left in the millimeter scale reactor after the reactions and the micro-hydrogen generators have limited fuel storage space. The SEM images of the reaction byproducts indicates that the byproducts of LiBH 4 can form a single solid mass that clogged the reaction vessel and limit the full utilization of the hydride. However, the byproducts of LiAlH 4 and CaH 2 reactions are non-agglomerated and they do not form impermeable mass. The experimental results show that the highest yield of hydrogen generation was achieved with LiAlH 4 and CaH 2 fuels.

  15. REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND TOWARD NORTH INTO LOWER GRID CASTING. HOLES OF VARIOUS SIZES ACCOMMODATE COOLANT WATER AND EXPERIMENTAL POSITIONS. INL NEGATIVE NO. 4197. Unknown Photographer, 2/11/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  16. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    SciTech Connect

    Carvo, Alan E.

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  17. Optimization of hydraulic cement admixture waste forms for sodium-bearing, high aluminum, and high zirconium wastes

    SciTech Connect

    Herbst, A.K.

    1997-08-01

    A three-way blend of portland cement, blast furnace slag, and fly ash was successfully tested on simulated acidic high sodium, aluminum, and zirconium low-level wastes (LLW). Grout cubes were prepared at various waste loadings to maximize loading while meeting compressive strength and leach resistance requirements. For sodium LLW, a 21% waste loading achieves a volume reduction of 3.3 and a compressive strength of 2750 pounds per square inch while meeting leach, mix, and flow requirements. It was found that the sulfur in the slag reduces the chromium leach rate below regulatory limits. For aluminum LLW, a 10% waste loading achieves a volume reduction of 8.5 and a compressive strength of 4.50 pounds per square inch while meeting leach requirements. Likewise for zirconium LLW, a 21% waste loading achieves a volume reduction of 8.3 and a compressive strength of 3570 pounds per square inch.

  18. Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

    NASA Technical Reports Server (NTRS)

    Mireles, Omar R.; Houts, Michael G.

    2011-01-01

    Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

  19. Modelling of hydride cracking

    SciTech Connect

    Zheng, X.J.; Metzger, D.R.; Glinka, G.; Dubey, R.N.

    1996-12-01

    Zirconium alloys may be susceptible to hydride formation under certain service conditions, due to hydrogen diffusion and precipitation in the presence of stress concentrations and temperature gradients. The inhomogeneous brittle hydride platelets that form are modeled as plane defects of zero thickness, with fracture toughness less than that of the matrix. A fracture criterion based on sufficient energy and stress is proposed for either delayed hydride cracking (DHC) under constant loading conditions, or hydride cracking at rising loads, such as in a fracture toughness test. The fracture criterion is validated against available experimental data concerning initiation of hydride fracture in smooth specimens, and DHC in cracked specimens under various loading and temperature conditions.

  20. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  1. Hydride precipitation in titanium

    SciTech Connect

    Numakura, H.; Kowia, M.

    1984-10-01

    The crystal structure and morphology of hydride (deuteride) precipitates are investigated on ..cap alpha..-titanium specimens containing 1-3 at.% H or D by transmission electron microscopy. The hydride is found to have a face-centered tetragonal structure (c/a = 1.09) with an ordered arrangement of hydrogen, being isomorphous to ..gamma..-zirconium hydride. Two types of precipitation mode are observed with the habit planes (0110) and near (0225).

  2. Sodium Zirconium Cyclosilicate (ZS-9): A Novel Agent for the Treatment of Hyperkalemia.

    PubMed

    Linder, Kristin E; Krawczynski, Michelle A; Laskey, Dayne

    2016-08-01

    Hyperkalemia is a potentially life-threatening electrolyte abnormality that may be caused by select medications, underlying organ dysfunction, or alterations in potassium homeostasis. Treatment for this condition has remained largely unchanged since the release of sodium polystyrene sulfonate (SPS) in 1958. Despite its widespread use, the safety and efficacy of SPS remains controversial. Two novel potassium-binding resins have emerged in recent years. Patiromer was the first of these to receive U.S. Food and Drug Administration approval for the treatment of hyperkalemia in October 2015. A second potassium-binding resin, a zirconium cyclosilicate currently known as ZS-9, may provide yet another alternative to the archetypal treatment with SPS. ZS-9 is an orally administered nonabsorbed inorganic compound that selectively binds potassium ions in vivo. Two phase III multicenter, randomized, placebo-controlled, double-blind trials have evaluated ZS-9 for the treatment of acute hyperkalemia. In this review, we discuss the pharmacology, clinical efficacy, safety, and potential place in therapy of ZS-9 for the enhanced elimination of potassium in the setting of hyperkalemia. PMID:27393581

  3. Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

    SciTech Connect

    Sackett, J.I.

    1983-01-01

    This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance.

  4. Computational fluid dynamics-aided analysis of a hydride vapor phase epitaxy reactor

    NASA Astrophysics Data System (ADS)

    Schulte, Kevin L.; Simon, John; Roy, Abhra; Reedy, Robert C.; Young, David L.; Kuech, Thomas F.; Ptak, Aaron J.

    2016-01-01

    We report the development of a computational fluid dynamics (CFD) model of a dual chamber hydride vapor phase epitaxial (HVPE) growth reactor. Uniformity of reactant concentrations in the growth stream, transient reactor flows, and cross doping between the two growth chambers, all factors critical to the deposition of uniform, low defect semiconductor layers, were modeled. Simulation results were generated by solving the fundamental continuity, momentum and energy equations over a discretized reactor volume by a finite volume analysis with the aid of CFD-ACE+ commercial software. We demonstrated uniformity of the vapor composition within ±1% across the substrate, achieved due to specific features of the reactor design. Small compositional non-uniformity (±2% absolute) in In1-xGaxP layers grown in our reactor was correlated with calculated temperature non-uniformity across the substrate. Gas switching was modeled and the transient time predicted by the model was confirmed by measurement of doping transients in a sample grown in the reactor. Lastly the gas curtains that chemically isolate the reactor chambers were modeled and the results were compared to experimental data for cross doping between the chambers. As an example, we demonstrate, based on insight from the model, that our HVPE reactor is suitable for the deposition of GaAs PV devices. CFD modeling is a critical tool for the scale up of laboratory level processes to industrial levels.

  5. Sodium fast reactor fuels and materials : research needs.

    SciTech Connect

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  6. SPACE-R thermionic space nuclear power system: Design and technology demonstration. Task 1.5.6, Moderator containment laboratory experiment test plant (CDRL No. 5)

    SciTech Connect

    Not Available

    1993-10-01

    The preferred moderator being considered for SPACE-R is yttrium hydride encased in beryllium tubes. The baseline beryllium performs a dual function as it acts as a moderator and provides containment for hydrogen. The permeation rate of hydrogen from the hydride through the beryllium shell at the operating temperature is an important factor for the functionality and reliability of the Be-YHx moderator. Hydrogen containment capability of beryllium is comparable to enamel which was used in SNAP and Topaz II reactors. However, limited experimental data base exists for the hydrogen permeation through fabricated beryllium enclosures at high temperature. Permeation of hydrogen in beryllium is strongly affected by surface conditions, thickness of surface oxide, surface and bulk traps, impurity content and microstructure. The objective of this experiment is to determine the permeation rate of hydrogen from yttrium hydride and zirconium hydride through beryllium in the temperature range of 773 K--973 K. In addition, Topaz II type zirconium hydride specimens with and without the proprietary oxide coating canned in stainless steel will be tested to measure the hydrogen permeation rate. The TSET SS-canned ZrHx samples currently at Phillips Laboratory will be used for the latter test with Phillips Laboratory participation at the SPI hydrogen leak test stand. A key technology demonstration of the effectiveness of transferred arc plasma spraying of a 1 mil Molybdenum coating on the Be cladding will be performed. The effectiveness of the Molybdenum coating in preventing any interaction of Be with Stainless Steel in NaK will be assessed and demonstrated.

  7. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  8. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  9. Convective cooling in a pool-type research reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  10. Example Work Domain Analysis for a Reference Sodium Fast Reactor

    SciTech Connect

    Hugo, Jacques; Oxstrand, Johanna

    2015-01-01

    The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.

  11. Hydride vapor phase epitaxy of AlN using a high temperature hot-wall reactor

    NASA Astrophysics Data System (ADS)

    Baker, Troy; Mayo, Ashley; Veisi, Zeinab; Lu, Peng; Schmitt, Jason

    2014-10-01

    Aluminum nitride (AlN) was grown on c-plane sapphire substrates by hydride vapor phase epitaxy (HVPE). The experiments utilized a two zone inductively heated hot-wall reactor. The surface morphology, crystal quality, and growth rate were investigated as a function of growth temperature in the range of 1450-1575 °C. AlN templates grown to a thickness of 1 μm were optimized with double axis X-ray diffraction (XRD) rocking curve full width half maximums (FWHMs) of 135″ for the (002) and 513″ for the (102).

  12. Liquid Metal Cooled Reactor for Space Power

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  13. METHOD OF FABRICATING A GRAPHITE MODERATED REACTOR

    DOEpatents

    Kratz, H.R.

    1963-05-01

    S>A nuclear reactor formed of spaced bodies of uranium and graphite blocks is improved by diffusing helium through the graphite blocks in order to replace the air in the pores of the graphite with helium. The helium-impregnated graphite conducts heat better, and absorbs neutrons less, than the original air- impregnated graphite. (AEC)

  14. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  15. TRITIUM IN-BED ACCOUNTABILITY FOR A PASSIVELY COOLED, ELECTRICALLY HEATED HYDRIDE BED

    SciTech Connect

    Klein, J.; Foster, P.

    2011-01-21

    A PAssively Cooled, Electrically heated hydride (PACE) Bed has been deployed into tritium service in the Savannah River Site (SRS) Tritium Facilities. The bed design, absorption and desorption performance, and cold (non-radioactive) in-bed accountability (IBA) results have been reported previously. Six PACE Beds were fitted with instrumentation to perform the steady-state, flowing gas calorimetric inventory method. An IBA inventory calibration curve, flowing gas temperature rise ({Delta}T) versus simulated or actual tritium loading, was generated for each bed. Results for non-radioactive ('cold') tests using the internal electric heaters and tritium calibration results are presented. Changes in vacuum jacket pressure significantly impact measured IBA {Delta}T values. Higher jacket pressures produce lower IBA {Delta}T values which underestimate bed tritium inventories. The exhaust pressure of the IBA gas flow through the bed's U-tube has little influence on measured IBA {Delta}T values, but larger gas flows reduce the time to reach steady-state conditions and produce smaller tritium measurement uncertainties.

  16. Numerical comparison of hydrogen desorption behaviors of metal hydride beds based on uranium and on zirconium-cobalt

    SciTech Connect

    Kyoung, S.; Yoo, H.; Ju, H.

    2015-03-15

    In this paper, the hydrogen delivery capabilities of uranium (U) and zirconium-cobalt (ZrCo) are compared quantitatively in order to find the optimum getter materials for tritium storage. A three-dimensional hydrogen desorption model is applied to two identically designed cylindrical beds with the different materials, and hydrogen desorption simulations are then conducted. The simulation results show superior hydrogen delivery performance and easier thermal management capability for the U bed. This detailed analysis of the hydrogen desorption behaviors of beds with U and ZrCo will help to identify the optimal bed material, bed design, and operating conditions for the storage and delivery system in ITER. (authors)

  17. CO2 activation through silylimido and silylamido zirconium hydrides supported on N-donor chelating SBA15 surface ligands.

    PubMed

    Pasha, Farhan Ahmad; Bendjeriou-Sedjerari, Anissa; Abou-Hamad, Edy; Huang, Kuo-Wei; Basset, Jean-Marie

    2016-02-11

    Density functional theory calculations and 2D 1H-13C HETCOR solid state NMR spectroscopy prove that CO2 can be used to probe, by its own reactivity, different types of N-donor surface ligands on SBA15-supported Zr(IV) hydrides: [≡(Si-O-)(≡Si-N=)[Zr]H] and [≡(Si-NH-)(≡Si-X-)[Zr]H2] (X=O or NH). Moreover, [≡(Si–O-)(≡Si-N=)[Zr]H] activates CO2 more efficiently than the other complexes and leads to the formation of a carbimato Zr formate. PMID:26750777

  18. LFR "Lead-Cooled Fast Reactor"

    SciTech Connect

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  19. Fundamental experiments on hydride reorientation in zircaloy

    NASA Astrophysics Data System (ADS)

    Colas, Kimberly B.

    In the current study, an in-situ X-ray diffraction technique using synchrotron radiation was used to follow directly the kinetics of hydride dissolution and precipitation during thermomechanical cycles. This technique was combined with conventional microscopy (optical, SEM and TEM) to gain an overall understanding of the process of hydride reorientation. Thus this part of the study emphasized the time-dependent nature of the process, studying large volume of hydrides in the material. In addition, a micro-diffraction technique was also used to study the spatial distribution of hydrides near stress concentrations. This part of the study emphasized the spatial variation of hydride characteristics such as strain and morphology. Hydrided samples in the shape of tensile dog-bones were used in the time-dependent part of the study. Compact tension specimens were used during the spatial dependence part of the study. The hydride elastic strains from peak shift and size and strain broadening were studied as a function of time for precipitating hydrides. The hydrides precipitate in a very compressed state of stress, as measured by the shift in lattice spacing. As precipitation proceeds the average shift decreases, indicating average stress is reduced, likely due to plastic deformation and morphology changes. When nucleation ends the hydrides follow the zirconium matrix thermal contraction. When stress is applied below the threshold stress for reorientation, hydrides first nucleate in a very compressed state similar to that of unstressed hydrides. After reducing the average strain similarly to unstressed hydrides, the average hydride strain reaches a constant value during cool-down to room temperature. This could be due to a greater ease of deforming the matrix due to the applied far-field strain which would compensate for the strains due to thermal contraction. Finally when hydrides reorient, the average hydride strains become tensile during the first precipitation regime and

  20. Development and application of modeling tools for sodium fast reactor inspection

    SciTech Connect

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-18

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  1. Development and application of modeling tools for sodium fast reactor inspection

    NASA Astrophysics Data System (ADS)

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-01

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  2. Progress in reliability of fast reactor operation and new trends to increased inherent safety

    SciTech Connect

    Merk, Bruno; Stanculescu, Alexander; Chellapandi, Perumal; Hill, Robert

    2015-06-01

    The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed.

  3. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  4. EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES

    SciTech Connect

    Douglas L. Porter

    2011-02-01

    Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

  5. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    NASA Astrophysics Data System (ADS)

    Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

    2009-07-01

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  6. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  7. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  8. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    SciTech Connect

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  9. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

  10. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

    Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  11. Insertion and [beta]-hydride elimination reactions of ruthenium/zirconium complexes containing C[sub 2] bridges with bond orders of 1, 2, and 3

    SciTech Connect

    Lemke, F.R.; Bullock, R.M. )

    1992-12-01

    Carbon dioxide inserts into the Zr-C bond of Cp(PMe[sub 3])[sub 2]RuCH[double bond]CHZrClCp[sub 2], producing the carboxylate complex Cp(PMe[sub 3])[sub 2]RuCH[double bond]CHCO[sub 2]ZrClCp[sub 2]. An [eta][sup 2]-iminoacyl insertion product, Cp(PMe[sub 3])[sub 2]Ru-C[triple bond]C-C(N[sub t]Bu)ZrClCp[sub 2], results from insertion of [sup t]BuNC into the Zr-C bond of the dimetalloalkyne complex Cp(PMe[sub 3])[sub 2]Ru-C[triple bond]C-ZrClCp[sub 2]. The reaction of Cp(PMe[sub 3])[sub 2]RuCH[sub 2]CH[sub 2]ZrClCp[sub 2] with high concentrations of [sup t]BuNC also produces an [eta][sup 2]-iminoacyl insertion product, Cp(PMe[sub 3])[sub 2]RuCH[sub 2]CH[sub 2]C(N[sup t]Bu)ZrClCp[sub 2]. At low concentrations of [sup t]BuNC, a competing reaction is [beta]-hydride elimination from Cp(PMe[sub 3])[sub 2]RuCH[sub 2]CH[sub 2]ZrClCp[sub 2] to give Cp(PMe[sub 3])[sub 2]RuCH[double bond]CH[sub 2] and Cp[sub 2]Zr(H)Cl, which is trapped by [sup t]BuNC under these conditions to give Cp[sub 2]ZrCl([eta][sup 2]-[sup t]BuN[double bond]CH). The reaction of benzophenone with Cp(PMe[sub 3])[sub 2]RuCH[sub 2]CH[sub 2]ZrClCp[sub 2] gives the insertion product Cp(PMe[sub 3])[sub 2]RuCH[sub 2]CH[sub 2]CPh[sub 2]OZrClCp[sub 2]. [beta]-Hydride elimination is also observed from Cp(PMe[sub 3])[sub 2]RuCH[double bond]CHZrClCp[sub 2]; in the presence of excess [sup n]BuC[triple bond]CH, the products are Cp(PMe[sub 3])[sub 2]Ru-C[triple bond]C-H and the zirconium vinyl complex Cp[sub 2]Zr(Cl)CH[double bond]CH[sup n]Bu. 25 refs.

  12. Sub-Doppler cooling of sodium atoms in gray molasses

    NASA Astrophysics Data System (ADS)

    Colzi, Giacomo; Durastante, Gianmaria; Fava, Eleonora; Serafini, Simone; Lamporesi, Giacomo; Ferrari, Gabriele

    2016-02-01

    We report on the realization of sub-Doppler laser cooling of sodium atoms in gray molasses using the D1 optical transition (3 s 1/2 2S →3 p 1/2 2P) at 589.8 nm. The technique is applied to samples containing 3 ×109 atoms, previously cooled to 350 μ K in a magneto-optical trap, and it leads to temperatures as low as 9 μ K and phase-space densities in the range of 10-4. The capture efficiency of the gray molasses is larger than 2/3, and we observe no density-dependent heating for densities up to 1011cm-3 .

  13. Ductility Evaluation of As-Hydrided and Hydride Reoriented Zircaloy-4 Cladding under Simulated Dry-Storage Condition

    SciTech Connect

    Yan, Yong; Plummer, Lee K; Ray, Holly B; Cook, Tyler S; Bilheux, Hassina Z

    2014-01-01

    Pre-storage drying-transfer operations and early stage storage expose cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to normal operation in-reactor and pool storage under these conditions. Radial hydrides could precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature. As a means of simulating this behavior, unirradiated hydrided Zircaloy-4 samples were fabricated by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. Mechanical testing was carried out by the ring compression test (RCT) method at various temperatures to evaluate the sample s ductility for both as-hydrided and post-hydride reorientation treated specimens. As-hydrided samples with higher hydrogen concentration (>800 ppm) resulted in lower strain before fracture and reduced maximum load. Increasing RCT temperatures resulted in increased ductility of the as-hydrided cladding. A systematic radial hydride treatment was conducted at various pressures and temperatures for the hydrided samples with H content around 200 ppm. Following the radial hydride treatment, RCTs on the hydride reoriented samples were conducted and exhibited lower ductility compared to as-hydrided samples.

  14. Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

    SciTech Connect

    Graf, A.; Petrick, H.; Stutz, U.; Hosking, P.

    2012-07-01

    The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and

  15. High Temperature Gas-Cooled Test Reactor Options Status Report

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  16. Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1

    SciTech Connect

    Dunn, F.E.

    1987-01-01

    The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.

  17. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  18. Determination of inorganic arsenic species by flow injection hydride generation atomic absorption spectrometry with variable sodium tetrahydroborate concentrations*1

    NASA Astrophysics Data System (ADS)

    Sigrist, Mirna E.; Beldoménico, Horacio R.

    2004-07-01

    This work describes a study on the determination of inorganic arsenic species in ground water and synthetic experimental matrices, using a flow injection system with on-line hydride generation device coupled to an atomic absorption spectrometer with flame-heated quartz atomizer (FI HG AAS). Specific trivalent arsenic determination is based on the slow kinetics of As(V) on the hydride generation reaction using sufficiently low concentrations of sodium tetrahydroborate (NaBH 4) as reductant in highly acidic conditions (pH<0). Under these conditions, the efficiency of hydride generation from As(V) is much lower than that from As(III). The pentavalent form is determined by the difference between total inorganic arsenic and As(III). As(V) interferences were studied using As(III) solutions ranging from 0% to 50% of total inorganic As. The optimized NaBH 4 concentration was 0.035% (w/v). The detection limit was 1.4 μg l -1 As(III). As(V) interferences were 6% in the case of water samples with 6 μg l -1 As(III) in the presence of 54 μg l -1 As(V) (i.e. 10% As(III)). Interferences of methylated arsenic species (MMA and DMA) were evaluated. Speciation method was satisfactorily applied to 20 field arsenical water samples from Santa Fe, Argentina, with values ranging from 30 to 308 μg l -1 total As. We found from 0% to 36% As(III) in the 20 field samples. The developed methodology constitutes an economic, simple and reliable way to evaluate inorganic arsenic distribution in underground waters or similar systems with negligible or no content of organoarsenicals.

  19. Multidimensional simulations of hydrides during fuel rod lifecycle

    NASA Astrophysics Data System (ADS)

    Stafford, D. S.

    2015-11-01

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

  20. Hydride precipitation kinetics in Zircaloy-4 studied using synchrotron X-ray diffraction

    NASA Astrophysics Data System (ADS)

    Courty, Olivier F.; Motta, Arthur T.; Piotrowski, Christopher J.; Almer, Jonathan D.

    2015-06-01

    As a result of in-reactor corrosion during operation in nuclear reactors, hydrogen can enter the zirconium fuel cladding and precipitate as brittle hydride particles, which may reduce cladding ductility. Dissolved hydrogen responds to temperature gradients, resulting in transport and precipitation into cold spots so that the distribution of hydrides in the cladding is inhomogeneous. The hydrogen precipitation kinetics plays a strong role in the spatial distribution of the hydrides in the cladding. The precipitation rate is normally described as proportional to the supersaturation of hydrogen in solid solution. The proportionality constant, α2, for hydride precipitation in Zircaloy-4 is measured directly using in situ synchrotron X-Ray diffraction, at different temperatures and with three different initial hydrogen concentrations. The results validate the linear approximation of the phenomenological model and a near constant value of α2 = 4.5 × 10-4 s-1 was determined for the temperature range studied.

  1. Hydride precipitation kinetics in Zircaloy-4 studied using synchrotron X-ray diffraction

    SciTech Connect

    Courty, Olivier Fabrice; Motta, Arthur T.; Piotrowski, Christopher J.; Almer, Jonathan D.

    2015-01-01

    As a result of in-reactor corrosion during operation in nuclear reactors, hydrogen can enter the zirconium fuel cladding and precipitate as brittle hydride particles, which may reduce cladding ductility. Dissolved hydrogen responds to temperature gradients, resulting in transport and precipitation into cold spots so that the distribution of hydrides in the cladding is inhomogeneous. The hydrogen precipitation kinetics plays a strong role in the spatial distribution of the hydrides in the cladding. The precipitation rate is normally described as proportional to the supersaturation of hydrogen in solid solution. The proportionality constant, α2, for hydride precipitation in Zircaloy-4 is measured directly using in situ synchrotron X-Ray diffraction, at different temperatures and with three different initial hydrogen concentrations. The results validate the linear approximation of the phenomenological model and a near constant value of α2 = 4.5 × 10-4 s-1 was determined for the temperature range studied.

  2. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  3. Code System for Supercritical Water Cooled Reactor LOCA Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1999-10-13

    Version 00 The new SCRELA code was developed to analyze the LOCA of the supercritical water cooled reactor. Since the currently available LWR codes for LOCA analysis could not analyze the significant differences in reactor characteristics between the supercritical-water cooled reactor and the current LWR, the first objective of this code development was to analyze the uniqueness of this reactor. The behavior of the supercritical water in the blowdown phase and the reflood phase ismore » modeled.« less

  4. Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Massacret, N.; Moysan, J.; Ploix, M. A.; Jeannot, J. P.; Corneloup, G.

    2013-01-01

    In the framework of the French R&D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 °C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlabin order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

  5. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  6. Comparison of sodium zirconium phosphate-structured HLW forms and synroc for high-level nuclear waste immobilization

    SciTech Connect

    Zyryanov, V.N.; Vance, E.R.

    1996-12-31

    The incorporation of (a) Cs/Sr as simulated heat-generating isotopes contained in Purex reprocessing waste, (b) simulated actinides, and (c) simulated Purex waste in sodium zirconium phosphate (NZP) has been studied. The samples were prepared by sintering, by hot pressing and by hot isostatic pressing in metal bellows containers. The short-term chemical durability of the phosphate-based material containing Purex waste was within an order of magnitude of that for Synroc-C, as measured by 7-day MCC-1 tests at 90{degrees}C. The dissolution behavior showed evidence of re-precipitation phenomena, even after times as short as 28 days. Potential for improvement of NZP-based ceramics for HLW management is discussed. 19 refs., 4 figs., 3 tabs.

  7. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    SciTech Connect

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver; J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  8. Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

    SciTech Connect

    Benoit, J. C.; Bourdot, P.; Eschbach, R.; Boucher, L.; Pascal, V.; Fontaine, B.; Martin, L.; Serot, O.

    2012-07-01

    A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

  9. MHTGR (Modular High-Temperature Gas-Cooled Reactor) design and development status

    SciTech Connect

    Turner, R.F.; Neylan, A.J.

    1988-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced power plant concept which has been under design definition since 1984. The design utilizes basic high-temperature gas-cooled reactor features of ceramic fuel, helium coolant and a graphite moderator which have been under development for 30 years. The geometric arrangement of the reactor vessels, the core and the heat removal components has been selected to exploit the inherent characteristics associated with high temperature materials. The design utilizes passively safe features which provide a higher margin of safety and investment protection than current generation reactors. The design has been evaluated to be economically attractive relative to modern coal fired plants. The design and development program is a cooperative effort by the US government, the utilities and the nuclear industry. 8 refs., 4 figs., 4 tabs.

  10. Limiting Factors for External Reactor Vessel Cooling

    SciTech Connect

    Cheung, F.B.

    2005-11-15

    The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

  11. Preliminary engineering design of sodium-cooled CANDLE core

    SciTech Connect

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-06

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CANDLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  12. Preliminary engineering design of sodium-cooled CANDLE core

    NASA Astrophysics Data System (ADS)

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-01

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  13. Particle bed reactor nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Ludewig, Hans

    1991-01-01

    The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times.

  14. Feasibility of Water Cooled Thorium Breeder Reactor Based on LWR Technology

    SciTech Connect

    Takaki, Naoyuki; Permana, Sidik; Sekimoto, Hiroshi

    2007-07-01

    The feasibility of Th-{sup 233}U fueled, homogenous breeder reactor based on matured conventional LWR technology was studied. The famous demonstration at Shipping-port showed that the Th-{sup 233}U fueled, heterogeneous PWR with four different lattice fuels was possible to breed fissile but its low averaged burn-up including blanket fuel and the complicated core configuration were not suitable for economically competitive reactor. The authors investigated the wide design range in terms of fuel cell design, power density, averaged discharge burn-up, etc. to determine the potential of water-cooled Th reactor as a competitive breeder. It is found that a low moderated (MFR=0.3) H{sub 2}O-cooled reactor with comparable burn-up with current LWR is feasible to breed fissile fuel but the core size is too large to be economical because of the low pellet power density. On the other hand, D{sub 2}O-cooled reactor shows relatively wider feasible design window, therefore it is possible to design a core having better neutronic and economic performance than H{sub 2}O-cooled. Both coolant-type cores show negative void reactivity coefficient while achieving breeding capability which is a distinguished characteristics of thorium based fuel breeder reactor. (authors)

  15. Alternative cooling resource for removing the residual heat of reactor

    SciTech Connect

    Park, H. C.; Lee, J. H.; Lee, D. S.; Jung, C. Y.; Choi, K. Y.

    2012-07-01

    The Recirculated Cooling Water (RCW) system of a Candu reactor is a closed cooling system which delivers demineralized water to coolers and components in the Service Building, the Reactor Building, and the Turbine Building and the recirculated cooling water is designed to be cooled by the Raw Service Water (RSW). During the period of scheduled outage, the RCW system provides cooling water to the heat exchangers of the Shutdown Cooling System (SDCS) in order to remove the residual heat of the reactor, so the RCW heat exchangers have to operate at all times. This makes it very hard to replace the inlet and outlet valves of the RCW heat exchangers because the replacement work requires the isolation of the RCW. A task force was formed to prepare a plan to substitute the recirculated water with the chilled water system in order to cool the SDCS heat exchangers. A verification test conducted in 2007 proved that alternative cooling was possible for the removal of the residual heat of the reactor and in 2008 the replacement of inlet and outlet valves of the RCW heat exchangers for both Wolsong unit 3 and 4 were successfully completed. (authors)

  16. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  17. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  18. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  19. The oxidation and hydriding of zircaloy fuel cladding in high temperature aqueous solutions

    NASA Astrophysics Data System (ADS)

    Chen, Yingzi

    Nearly 90% of today's fission reactors use Zr based fuel cladding materials. The Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) are the two most common water-cooled nuclear reactors. Corrosion is the principal threat to the failure of the fuel in these reactors, resulting in the release of fission products to the coolant and hence to the establishment of radiation fields in out-of-core regions of the coolant circuit (e.g., steam generators in PWRs and turbines in BWRs). As is well known, corrosion is an electrochemical phenomenon; however, electrochemical effects are often neglected in corrosion studies on zirconium and its alloys, because of the difficulty in performing well-defined experiments under the appropriate conditions (high temperatures and pressures). In-situ studies have been carried out to examine the electrochemistry of passive zirconium under simulated BWR and PWR coolant conditions by using a controlled hydrodynamic, high temperature/high pressure test cell. The oxidation/hydriding mechanisms are elucidated by measuring the current, impedance, and capacitance of passive zirconium as a function of formation potential. The data are interpreted in terms of a modified point defect model (PDM) that recognize the existence of a passive film comprising a thick oxide outer layer over a thin barrier layer. From the composition of the zirconium passive film and thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Transients in current density and the thickness of the passive film formed on zirconium, when stepping the potential in either the positive or negative directions, have confirmed that the rate law afforded by the PDM adequately describes the growth and thinning of the passive film at high temperatures. The experimental results demonstrate that the kinetics of either oxygen or hydrogen vacancy generation

  20. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    DOEpatents

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  1. U.S./CIS eye joint nuclear rocket venture

    NASA Technical Reports Server (NTRS)

    Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

    1993-01-01

    An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

  2. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  3. The reactivity of sodium alanates with O[2], H[2]O, and CO[2] : an investigation of complex metal hydride contamination in the context of automotive systems.

    SciTech Connect

    Dedrick, Daniel E.; Bradshaw, Robert W.; Behrens, Richard, Jr.

    2007-08-01

    Safe and efficient hydrogen storage is a significant challenge inhibiting the use of hydrogen as a primary energy carrier. Although energy storage performance properties are critical to the success of solid-state hydrogen storage systems, operator and user safety is of highest importance when designing and implementing consumer products. As researchers are now integrating high energy density solid materials into hydrogen storage systems, quantification of the hazards associated with the operation and handling of these materials becomes imperative. The experimental effort presented in this paper focuses on identifying the hazards associated with producing, storing, and handling sodium alanates, and thus allowing for the development and implementation of hazard mitigation procedures. The chemical changes of sodium alanates associated with exposure to oxygen and water vapor have been characterized by thermal decomposition analysis using simultaneous thermogravimetric modulated beam mass spectrometry (STMBMS) and X-ray diffraction methods. Partial oxidation of sodium alanates, an alkali metal complex hydride, results in destabilization of the remaining hydrogen-containing material. At temperatures below 70 C, reaction of sodium alanate with water generates potentially combustible mixtures of H{sub 2} and O{sub 2}. In addition to identifying the reaction hazards associated with the oxidation of alkali-metal containing complex hydrides, potential treatment methods are identified that chemically stabilize the oxidized material and reduce the hazard associated with handling the contaminated metal hydrides.

  4. Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

    SciTech Connect

    Conklin, J.C. )

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

  5. Moderate temperature sodium cells. I - Transition metal disulfide cathodes

    NASA Technical Reports Server (NTRS)

    Abraham, K. M.; Pitts, L.; Schiff, R.

    1980-01-01

    TiS2, VS2, and Nb(1.1)S2 transition metal disulfides were evaluated as cathode materials for a moderate temperature rechargeable Na cell operating at 130 C. The 1st discharge of TiS2 results in a capacity of 0.85 eq/mole; approximately half of the Na in the 1st phase spanning the Na range from zero to 0.30 and almost all the Na in the 2nd phase spanning the 0.37 to 0.80 range are rechargeable. VS2 intercalates up to one mole of Na/mole of VS2 in the 1st discharge; the resulting Na(x)VS2 ternary consists of 3 phases in the 3 ranges of Na from zero to 1. Niobium disulfide undergoes a phase change in the 1st discharge; the average rechargeable capacity in extended cycling of this cathode is 0.50 eq/mole.

  6. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    NASA Astrophysics Data System (ADS)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  7. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    SciTech Connect

    Afifah, Maryam Su’ud, Zaki; Miura, Ryosuke; Takaki, Naoyuki; Sekimoto, H.

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  8. Studies of Polonium Removal from Molten Lead-Bismuth for Lead-Alloy-Cooled Reactor Applications

    SciTech Connect

    Jacopo Buongiorno; Ken Czerwinski; Eric Loewen; Chris Larson

    2004-09-01

    The isotope 210Po is the main product of neutron activation in fast reactors cooled by molten lead-bismuth eutectic (LBE). The isotope 210Po is a pure alpha emitter with a half-life of 138.38 days. For typical values of the neutron flux the 210Po concentration in the coolant can reach 1-10 Ci/kg. While exposure of plant personnel to Po is prevented under normal operating conditions because the primary system is sealed, Po does pose a radiological hazard during maintenance activities for which access to submerged structures is required as well as during accidents resulting in breach of the primary-system barrier. Obviously, continuous removal of Po from the LBE reduces this hazard. Therefore, it is important to understand the mechanisms by which Po is formed in and released from the LBE. We summarize research performed at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology to investigate the basic chemistry of four mechanisms of Po release, which could serve as the basis for a coolant cleanup system in LBE-cooled reactors. The mechanisms explored are lead polonide evaporation, formation of polonium hydride, rare-earth filtering, and alkaline extraction. For the key chemical species involved expressions are given for useful quantities such as formation energy, release, and deposition rates. It is concluded that the most promising removal mechanism is alkaline extraction, although a more systematic investigation of this mechanism is needed.

  9. Studies of Polonium Removal from Molten Lead-Bismuth for Lead-Alloy-Cooled Reactor Applications

    SciTech Connect

    Buongiorno, Jacopo; Loewen, Eric P.; Czerwinski, Kenneth; Larson, Christopher

    2004-09-15

    The isotope {sup 210}Po is the main product of neutron activation in fast reactors cooled by molten lead-bismuth eutectic (LBE). The isotope {sup 210}Po is a pure alpha emitter with a half-life of 138.38 days. For typical values of the neutron flux the {sup 210}Po concentration in the coolant can reach 1-10 Ci/kg. While exposure of plant personnel to Po is prevented under normal operating conditions because the primary system is sealed, Po does pose a radiological hazard during maintenance activities for which access to submerged structures is required as well as during accidents resulting in breach of the primary-system barrier. Obviously, continuous removal of Po from the LBE reduces this hazard. Therefore, it is important to understand the mechanisms by which Po is formed in and released from the LBE. We summarize research performed at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology to investigate the basic chemistry of four mechanisms of Po release, which could serve as the basis for a coolant cleanup system in LBE-cooled reactors. The mechanisms explored are lead polonide evaporation, formation of polonium hydride, rare-earth filtering, and alkaline extraction. For the key chemical species involved expressions are given for useful quantities such as formation energy, release, and deposition rates. It is concluded that the most promising removal mechanism is alkaline extraction, although a more systematic investigation of this mechanism is needed.

  10. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    SciTech Connect

    Merk, B.; Weiss, F. P.

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  11. Draft of M2 Report on Integration of the Hybrid Hydride Model into INL's MBM Framework for Review

    SciTech Connect

    Tikare, Veena; Weck, Philippe F.; Schultz, Peter A.; Clark, Blythe; Glazoff, Michael; Homer, Eric

    2014-07-01

    This report documents the development, demonstration and validation of a mesoscale, microstructural evolution model for simulation of zirconium hydride {delta}-ZrH{sub 1.5} precipitation in the cladding of used nuclear fuels that may occur during long-term dry storage. While the Zr-based claddings are manufactured free of any hydrogen, they absorb hydrogen during service, in the reactor by a process commonly termed ‘hydrogen pick-up’. The precipitation and growth of zirconium hydrides during dry storage is one of the most likely fuel rod integrity failure mechanisms either by embrittlement or delayed hydride cracking of the cladding. While the phenomenon is well documented and identified as a potential key failure mechanism during long-term dry storage (NUREG/CR-7116), the ability to actually predict the formation of hydrides is poor. The model being documented in this work is a computational capability for the prediction of hydride formation in different claddings of used nuclear fuels. This work supports the Used Fuel Disposition Research and Development Campaign in assessing the structural engineering performance of the cladding during and after long-term dry storage. This document demonstrates a basic hydride precipitation model that is built on a recently developed hybrid Potts-phase field model that combines elements of Potts-Monte Carlo and the phase-field models. The model capabilities are demonstrated along with the incorporation of the starting microstructure, thermodynamics of the Zr-H system and the hydride formation mechanism.

  12. R and D program for core instrumentation improvements devoted for French sodium fast reactors

    SciTech Connect

    Jeannot, J. P.; Rodriguez, G.; Jammes, C.; Bernardin, B.; Portier, J. L.; Jadot, F.; Maire, S.; Verrier, D.; Loisy, F.; Prele, G.

    2011-07-01

    Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within the scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)

  13. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect

    Mardiansah, Deby; Takaki, Naoyuki

    2010-06-22

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  14. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  15. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  16. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    NASA Astrophysics Data System (ADS)

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E.; Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-01

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  17. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    SciTech Connect

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E. Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-15

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  18. Liquid metal reactor air cooling baffle

    DOEpatents

    Hunsbedt, A.

    1994-08-16

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat. 3 figs.

  19. Use of zirconium(IV) arsenophosphate columns for cation exchange separation of metal ions interfering in the spectrophotometric determination of uranium with sodium diethyl dithiocarbamate

    SciTech Connect

    Varshney, K.G.; Agrawal, S.; Anwar, S.; Varshney, K.

    1985-01-01

    A simple cation exchange method has been developed for the quantitative separation of uranium from some metal ions which generally interfere in its spectrophotometric determination using sodium diethyl dithiocarbamate as a reagent. The method requires only a single bed operation and enables a satisfactory (Error + or - separation of uranium (UO/sub 2/ (II)) up to 1080 ..mu..g from ten metal ions on a 2 g column of zirconium (IV) arsenophosphate cation exchanger in H(I) form.

  20. Reactor with natural convection backup cooling system

    SciTech Connect

    Koutz, S.L.; Cavallaro, L.; Kapich, D.D.

    1988-07-26

    A nuclear reactor is described comprising: a closed vessel containing a pool of secondary coolant; a reactor core located within the pool; an inlet conduit system for directing primary coolant to the lower end of the core; pumping means for effecting flow of the primary coolant into the inlet conduit system; an outlet conduit system located above the core for receiving coolant from the upper end of the core and carrying it to an exterior outlet pipe; and pressure reduction means for reducing fluid pressure in the inlet conduit system adjacent all of the one or more opening therein by locally increasing fluid velocity adjacent all of the one or more openings.

  1. Heat pipe cooled reactors for multi-kilowatt space power supplies

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  2. PBF Cooling Tower. View from highbay roof of Reactor Building ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  3. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  4. PLUTONIUM-ZIRCONIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  5. Emersion Testing of Phenix Reactor Components From Liquid Sodium

    SciTech Connect

    Baque, F.

    2002-07-01

    The life extension of the Phenix LMFR involved the inspection of reactor vessel internal structures: among other techniques, a visual inspection was performed of the above core structure, fuel assembly heads and upper components. To make this inspection possible, a partial draining of the main vessel from primary liquid sodium was carried out (sodium at 180 and argon cover at 150 ). The test program aimed at obtaining further knowledge on the process of wetting of sodium - as pure metal - on Phenix Plant assembly heads - made of stainless steel -, as well as on the internal structure welding, was carried out from November 1998 to January 1999. The main results were as follows: - the sodium meniscus measured during sodium lowering against the non-wet vertical structures reaches 10 mm in height. On wetted structures, it reaches only 5.3 mm. - when sodium level decreases, the process if very regular. However, re-flooding is carried out in stages. - a difference of 0.2 mm between two heads altitudes is enough to observe successively each of the heads. - the quality of sodium does not modify the wetting process (in the range of cold trap temperature: 110-140 deg. C). - the influence of lighting is important. - the visibility limit of emerging electro-eroded cracks (from 0.17 to 1.0 mm) is at 0.20 mm. - the visibility of a horizontal welding, machined or not, is good when the lighting is sufficient. - the superficial flow of sodium only modifies the wetting process for the closest heads. A final test allowed to observe that the global inclination of the assembly head mock-up does not modify the wetting process. These experimental results were part of the feasibility demonstration of the visual inspection within the actual Phenix Plant that was undertaken in 2001. (authors)

  6. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    SciTech Connect

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  7. Design of Recycle Pressurized Water Reactor with Heavy Water Moderation

    SciTech Connect

    Hibi, Koki; Uchita, Masato

    2004-03-15

    This study presents the conceptual design of the recycle pressurized water reactor (RPWR), which is an innovative PWR fueled with mixed oxide, moderated by heavy water, and having breeding ratios around 1.1. Most of the systems of RPWR can employ those of PWRs. The RPWR has no boric acid systems and has a small tritium removal system. The construction and operation costs would be similar to those of current PWRs. Heavy water cost has decreased drastically with up-to-date producing methods. The reliability of the systems of the RPWR is high, and the research and development cost for RPWR is very low because the core design is fundamentally based on the current PWR technology.

  8. Laser ablated zirconium plasma: A source of neutral zirconium

    SciTech Connect

    Yadav, Dheerendra; Thareja, Raj K.

    2010-10-15

    The authors report spectroscopic investigations of laser produced zirconium (Zr) plasma at moderate laser fluence. At low laser fluence the neutral zirconium species are observed to dominate over the higher species of zirconium. Laser induced fluorescence technique is used to study the velocity distribution of ground state neutral zirconium species. Two-dimensional time-resolved density distributions of ground state zirconium is mapped using planner laser induced fluorescence imaging and total ablated mass of neutral zirconium atoms is estimated. Temporal and spatial evolutions of electron density and temperature are discussed by measuring Stark broadened profile and ratio of intensity of emission lines, respectively.

  9. Emergency core cooling during an SRS reactor LOPA

    SciTech Connect

    Paul, P.K.; Burge, D.A.; Reeves, K.K.; Herman, D.T. )

    1993-01-01

    The loss-of-pumping accident (LOPA) is a Savannah River site (SRS) reactor design-basis accident. The most limiting LOPA is caused by a double-ended guillotine break in a secondary cooling system inlet header and is the topic of this discussion. Upon break detection, the reactor scrams and the secondary cooling water pumps and alternating-current (ac) primary pump motors trip off. The direct-current (dc) motors continue to drive the primary pumps at about one-third capacity. Gravity flow through the broken header continues flooding the building after the cooling pumps are off. The emergency cooling system (ECS) is activated prior to flood-out of the dc motors. The design-basis accident reactor power limit ensures the reactor will shut down safely should a LOPA occur. The simulated LOPA has five phases: steady state, ac coastdown, dc flow, dc coastdown, and fully developed ECS flow. Analyses of LOPAs have shown that ECS is the most limiting phase of the accident. This paper concentrates on the role of ECS in LOPA limits.

  10. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

  11. Fuel leak detection apparatus for gas cooled nuclear reactors

    DOEpatents

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  12. Small LBE-Cooled Fast Reactor for Expanding Market

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino; Kunihiko Nakamura; Yoshio Kamishima; Takashi Kawakita

    2002-07-01

    A long-life safe simple small portable proliferation-resistant reactor is expected to solve many problems associating future energy and globally environmental problems. From discussions on mainly neutronics and safety points it has been shown that the heavy liquid metal cooled fast reactor is the best candidate to satisfy the above requirements. A lead-bismuth-eutectic (LBE) cooled fast reactor LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor) has been designed and continues to be improved. In the present paper a recent version of LSPR is presented. The total power of the present design is 150 MWt (53 MWe). During whole reactor life of 12 years the excess reactivity required for burnup is very low, and negative coolant dilatation coefficient is confirmed. This characteristic together with some other characteristics makes unprotected loss of flow (ULOF) accident inherently safe. It can survive even simultaneous rod run-out transient over power (UTOP), ULOF and unprotected loss of heat sink (ULOHS) accident without the help of an operator or active device. (authors)

  13. Investigations on the phase equilibria of some hydride ion conducting electrolyte systems and their application for hydrogen monitoring in sodium coolant

    NASA Astrophysics Data System (ADS)

    Joseph, Kitheri; Sujatha, K.; Nagaraj, S.; Mahendran, K. H.; Sridharan, R.; Periaswami, G.; Gnanasekaran, T.

    2005-09-01

    Electrochemical meters for measuring hydrogen levels in liquid sodium need thermodynamically stable hydride ion conducting electrolytes. In order to identify electrolytes that have high hydride ion conductivity, phase diagram of systems consisting of low melting compounds such as CaCl 2-LiCl, SrBr 2-LiBr, SrBr 2-SrHBr and CaBr 2-CaHBr were investigated by differential scanning calorimetry and their phase diagrams established. Using these information and supplementary information on effects of addition of alkaline earth hydride to these systems, potential electrolytes were tested for their use in electrochemical meters. Meters were constructed using electrolytes with (i) 22mol%SrCl 2-12.2mol%CaCl 2-54.5mol%LiCl-11.3mol%CaHCl, (ii) 70mol%LiCl-16mol%CaHCl-14mol%CaCl 2 and (iii) 40mol%CaHBr-60mol%CaBr 2 compositions. Output of meters that had Li ions in liquid phase electrolyte showed non-linearity at low hydrogen levels. Output of meters using CaBr 2-40mol%CaHBr solid showed linearity in the concentration range of 50-250 ppb in sodium.

  14. Investigation of vessel exterior air cooling for a HLMC reactor

    SciTech Connect

    Sienicki, J. J.; Spencer, B. W.

    2000-01-13

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  15. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

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

    SciTech Connect

    Stout, R.B.

    1989-10-01

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

  17. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  18. Bottom shield for a gas cooled high temperature nuclear reactor

    SciTech Connect

    Schoening, J.; Elter, C.; Kolodzey, H. J.; Schwiers, H. G.; Stracke, W.

    1984-12-25

    A gas cooled, high temperature nuclear reactor is provided with a base plate arranged under the reactor core and over the bottom of the prestressed concrete pressure vessel serving as the bottom shield. The bottom shield comprises at least two plates arranged coaxially with respect to each other, one above the other. Each plate comprises several partially interconnected parts with the lower plate being placed at an axial and vertical distance from the bottom liner of the prestressed concrete pressure vessel and also from the upper plate.

  19. System Study: Reactor Core Isolation Cooling 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  20. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  1. System Study: Reactor Core Isolation Cooling 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  2. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

    SciTech Connect

    Steven R. Sherman

    2005-04-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

  3. Decay heat removal in GEN IV gas cooled fast reactors.

    SciTech Connect

    Cheng, L. Y.; Wei, T. Y. C.

    2009-08-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  4. Optimized evaporative cooling for sodium Bose-Einstein condensation against three-body loss

    SciTech Connect

    Shobu, Takahiko; Yamaoka, Hironobu; Imai, Hiromitsu; Morinaga, Atsuo; Yamashita, Makoto

    2011-09-15

    We report on a highly efficient evaporative cooling optimized experimentally. We successfully created sodium Bose-Einstein condensates with 6.4x10{sup 7} atoms starting from 6.6x10{sup 9} thermal atoms trapped in a magnetic trap by employing a fast linear sweep of radio frequency at the final stage of evaporative cooling so as to overcome the serious three-body losses. The experimental results such as the cooling trajectory and the condensate growth quantitatively agree with the numerical simulations of evaporative cooling on the basis of the kinetic theory of a Bose gas carefully taking into account our specific experimental conditions. We further discuss theoretically a possibility of producing large condensates, more than 10{sup 8} sodium atoms, by simply increasing the number of initial thermal trapped atoms and the corresponding optimization of evaporative cooling.

  5. The Case for Moderately-Cooled, Far-Infrared Thermal Detectors

    NASA Technical Reports Server (NTRS)

    Brasunas, John C.; Lakew, Brook

    2004-01-01

    There are moderately-cooled (around 77K) infrared detectors, for instance InSb (around 5 microns wavelength) and HgCdTe (around 15 to 20 microns wavelength). However for longer wavelengths there are either uncooled thermal-type detectors or highly cooled (about 4K and lower) quantum and thermal detectors, with the notable exception of high Tc superconductor detectors. We will describe certain long-wavelength applications in space where only moderate cooling is feasible, and where better sensitivity is required than possible with uncooled detectors. These requirements could be met with high Tc bolometers, but it may also be prudent to develop other technologies. Additionally, over the past 16 years a marketplace has not developed for the commercial production of high Tc bolometers, indicating their production may be a natural endeavor for government laboratories.

  6. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  7. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  8. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1{percent}Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. {copyright} {ital 1999 American Institute of Physics.}

  9. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-22

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  10. Investigation of heat and mass transfer process in metal hydride hydrogen storage reactors, suitable for a solar powered water pump system

    NASA Astrophysics Data System (ADS)

    Coldea, I.; Popeneciu, G.; Lupu, D.; Misan, I.; Blanita, G.; Ardelean, O.

    2012-02-01

    The paper analyzes heat and mass transfer process in metal hydride hydrogen storage systems as key element in the development of a solar powered pump system. Hydrogen storage and compression performance of the developed reactors are investigated according to the type of metal alloys, the metal hydride bed parameters and system operating conditions. To reach the desired goal, some metal hydride from groups AB5 and AB2 were synthesized and characterized using elements substitution for tailoring their properties: reversible hydrogen absorption capacity between the hydrogen absorption and desorption pressures at equilibrium at small temperature differences. For the designed hydrogen storage reactors, a new technical solution which combines the effective increase of the thermal conductivity of MH bed and good permeability to hydrogen gas circulation, was implemented and tested. The results permitted us to develop a heat engine with metal hydride, the main element of the functional model of a heat operated metal hydride based water pumping system using solar energy. This is a free energy system able to deliver water, at a convenience flow and pressure, in remote places without conventional energy access.

  11. Recent Sodium Technology Development for the Decommissioning of the Rapsodie and Superphenix Reactors and the Management of Sodium Wastes

    SciTech Connect

    Rodriguez, G.; Gastaldi, O.; Baque, F.

    2005-04-15

    The Commissariat a l'Energie Atomique (CEA) has recently developed and/or conducted experiments on several processes in support of the decommissioning of two French liquid-metal fast reactors (LMFRs), Rapsodie and Superphenix, as well as on the treatment of CEA sodium wastes. CEA has demonstrated that it is possible to define appropriate and efficient processes to meet the different situations encountered in decommissioning LMFRs. Mechanical techniques derived from standard technologies have been successfully applied to fast reactor decommissioning to complete primary vessel draining from sodium. In addition, specific chemical processes have been developed to deal safely with metallic sodium reactivity. Sodium-contaminated equipment has been successfully cleaned by reacting sodium with water mist in an atmosphere with carbon dioxide to form inert sodium carbonate. Bulk sodium has been successfully converted into aqueous caustic soda by injection of liquid-metallic sodium into sodium hydroxide solution. Several processes were also defined to deal with specific sodium wastes. In all cases the principle is based on a sodium/water chemical reaction where the released hydrogen and heat are controlled. With the development of a wide variety of processes, all steps in the decommissioning of LMFRs are assumed to be now properly mastered.

  12. Modification of the Core Cooling System of TRIGA 2000 Reactor

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-22

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  13. Modification of the Core Cooling System of TRIGA 2000 Reactor

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  14. High power density reactors based on direct cooled particle beds

    SciTech Connect

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs.

  15. Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap

    SciTech Connect

    Holcomb, David Eugene; Flanagan, George F; Mays, Gary T; Pointer, William David; Robb, Kevin R; Yoder Jr, Graydon L

    2014-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant; the lack of an approved licensing framework; the lack of qualified, salt-compatible structural materials; and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  16. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

  17. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  18. Fine-grained zirconium-base material

    DOEpatents

    Van Houten, G.R.

    1974-01-01

    A method is described for making zirconium with inhibited grain growth characteristics, by the process of vacuum melting the zirconium, adding 0.3 to 0.5% carbon, stirring, homogenizing, and cooling. (Official Gazette)

  19. Investigation of in-reactor creep and irradiation growth of zirconium-2.5 wt% niobium channel tubes

    NASA Astrophysics Data System (ADS)

    Goncharov, V. V.; Platonov, P. A.; Ivanov, A. N.; Frolov, I. A.

    1980-05-01

    We summarize the diametral creep results obtained in the MR reactor of the Kurchatov Institute of Atomic Energy on zirconium-2.5 wt% niobium pressure tubes of the type used in RBMK-1000 power reactors. The experiments that lasted up to 30 000 h cover a temperature range of 270 to 350°C, neutron fluxes between 0.6 and 4.0 ×10 13n/ cm2 · s ( E > 1 MeV) and stresses of up to 16 kgf/mm 2. Diametral strains of up to 4.8% have been measured. In-reactor creep results have been analyzed in terms of thermal and irradiation creep components assuming them to be additive. The thermal creep rate is given by a relationship of the type ɛth = A1exp [( A2 + A3 σt) T] and the irradiation component by ɛrad = A4 σtø( T - A5), where T = temperature, σ t = hoop stress, ø = neutron flux and a1 to A5 are constants. Irradiation growth experiments carried out at 280° C on specimens machined from pressure tubes showed a non-linear dependence of growth strain on neutron fluence up to neutron fluences of 5 × 10 20 n/cm 2. The significance of these results to the elongation of RBMK reactor pressure tubes is discussed.

  20. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

    SciTech Connect

    Steven R. Sherman; Collin J. Knight

    2006-03-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.

  1. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  2. Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

    SciTech Connect

    Massacret, N.; Jeannot, J. P.

    2013-01-25

    In the framework of the French R and D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 Degree-Sign C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlab Copyright-Sign in order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

  3. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    SciTech Connect

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  4. Scaling effects in sodium zirconium silicate phosphate (Na1+xZr2SixP3-xO12) ion-conducting thin films

    DOE PAGESBeta

    Ihlefeld, Jon F.; Gurniak, Emily; Jones, Brad H.; Wheeler, David R.; Rodriguez, Mark A.; McDaniel, Anthony H.

    2016-05-04

    Preparation of sodium zirconium silicate phosphate (NaSICon), Na1+xZr2SixP3–xO12 (0.25 ≤ x ≤ 1.0), thin films has been investigated via a chemical solution approach on platinized silicon substrates. Increasing the silicon content resulted in a reduction in the crystallite size and a reduction in the measured ionic conductivity. Processing temperature was also found to affect microstructure and ionic conductivity with higher processing temperatures resulting in larger crystallite sizes and higher ionic conductivities. The highest room temperature sodium ion conductivity was measured for an x = 0.25 composition at 2.3 × 10–5 S/cm. In conclusion, the decreasing ionic conductivity trends with increasingmore » silicon content and decreasing processing temperature are consistent with grain boundary and defect scattering of conducting ions.« less

  5. Brayton Cycle for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Oh, Chang H.; Moore, Richard L.

    2005-03-15

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others.The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  6. Brayton Cycle for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh

    2005-03-01

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others. The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  7. A helium-cooled blanket design of the low aspect ratio reactor

    SciTech Connect

    Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

    1998-03-01

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

  8. Reduction in cardiovascular risk by sodium-bicarbonated mineral water in moderately hypercholesterolemic young adults.

    PubMed

    Pérez-Granados, Ana M; Navas-Carretero, Santiago; Schoppen, Stefanie; Vaquero, M Pilar

    2010-10-01

    The effects of drinking sodium-bicarbonated mineral water on cardiovascular risk in young men and women with moderate cardiovascular risk were studied. Eighteen young volunteers (total cholesterol levels >5.2 mmol/L) without any disease participated. The study consisted of two 8-week intervention periods. Subjects consumed, as supplement to their usual diet, 1 L/day control low mineral water, followed by 1 L/day bicarbonated mineral water (48 mmol/L sodium, 35 mmol/L bicarbonate and 17 mmol/L chloride). Determinations were performed at the end of the control water period and on Weeks 4 and 8 of the bicarbonated water period. Body weight, body mass index (BMI), blood pressure, dietary intake, total cholesterol, low-density lipoprotein (LDL) cholesterol, high-density lipoprotein (HDL) cholesterol, apolipoprotein (Apo) A-I, Apo B, triacylgycerols, glucose, insulin, adiponectin, high-sensitivity C-reactive protein (hs-CRP), soluble adhesion molecules [soluble intercellular adhesion molecule (sICAM) and soluble vascular adhesion molecule (sVCAM)], sodium and chloride urinary excretion, and urine pH were measured. Dietary intake, body weight and BMI showed no significant variations. Systolic blood pressure decreased significantly after 4 weeks of bicarbonated water consumption, without significant differences between Weeks 4 and 8. After bicarbonated water consumption, significant reductions in total cholesterol (by 6.3%; P=.012), LDL cholesterol (by 10%; P=.001), total/HDL cholesterol (P=.004), LDL/HDL cholesterol (P=.001) and Apo B (P=.017) were observed. Serum triacylglycerol, Apo A-I, sICAM-1, sVCAM-1 and hs-CRP levels did not change. Serum glucose values tended to decrease during the bicarbonated water intervention (P=.056), but insulin levels did not vary. This sodium-bicarbonated mineral water improves lipid profile in moderately hypercholesterolemic young men and women and could therefore be applied in dietary interventions to reduce cardiovascular risk. PMID

  9. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  10. Selection of materials for sodium fast reactor steam generators

    SciTech Connect

    Dubiez-Le Goff, S.; Garnier, S.; Gelineau, O.; Dalle, F.; Blat-Yrieix, M.; Augem, J. M.

    2012-07-01

    Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)

  11. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  12. Cooling molten salt reactors using "gas-lift"

    NASA Astrophysics Data System (ADS)

    Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

    2014-08-01

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  13. Power Conversion Study for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gascooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language.

  14. Method of shielding a liquid-metal-cooled reactor

    DOEpatents

    Sayre, Robert K.

    1978-01-01

    The primary heat transport system of a nuclear reactor -- particularly for a liquid-metal-cooled fast-breeder reactor -- is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system.

  15. Cooling molten salt reactors using “gas-lift”

    SciTech Connect

    Zitek, Pavel E-mail: klimko@kke.zcu.cz; Valenta, Vaclav E-mail: klimko@kke.zcu.cz; Klimko, Marek E-mail: klimko@kke.zcu.cz

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  16. CFD modeling of high temperature gas cooled reactors

    SciTech Connect

    Janse van Rensburg, J.J.; Viljoen, C.; Van Staden, M.P.

    2006-07-01

    This paper presents an overview of how CFD has been applied to model the gas flow and heat transfer within the PBMR (Pebble Bed Modular reactor) with the aim of providing valuable design and safety information. The thermo-hydraulic calculations are performed using the STAR-CD [1] Computational Fluid Dynamics (CFD) code and the neutronic calculations are performed using VSOP [2]. Results are presented for steady-state normal operation and for a transient De-pressurized Loss Of Forced Cooling event (DLOFC). (authors)

  17. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  18. Monitoring system for a liquid-cooled nuclear fission reactor

    DOEpatents

    DeVolpi, Alexander

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  19. Neutronics investigation of advanced self-cooled liquid blanket systems in the helical reactor

    NASA Astrophysics Data System (ADS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M. Z.

    2008-03-01

    Neutronics investigations have been conducted in the design activity of the helical-type reactor Force Free Helical Reactor (FFHR2) adopting Flibe-cooled and Li-cooled advanced liquid blanket systems. In this study, comprehensive investigations and geometry modifications related to the tritium breeding ratios (TBRs), neutron shielding performance and neutron wall loading on the first walls in FFHR2 have been performed by improving the three-dimensional (3D) neutronics calculation system developed for non-axisymmetric helical designs. The total TBRs obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. However, it appeared that the most important neutronics issue in the present helical blanket configuration was suppression of neutron streaming through the divertor pumping areas and reflection from support structures for protection of poloidal and helical coils. Evaluation of neutron wall loading on the first walls indicated that the peaking factor would be moderated as low as 1.2 by the toroidal and helical effect of the helical-shaped plasma distribution in the helical reactor.

  20. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

    2013-07-01

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

  1. Pozzolanic filtration/solidification of radionuclides in nuclear reactor cooling water

    SciTech Connect

    Englehardt, J.D.; Peng, C.

    1995-12-31

    Laboratory studies to investigate the feasibility of one- and two-step processes for precipitation/coprecipitating radionuclides from nuclear reactor cooling water, filtering with pozzolanic filter aid, and solidifying, are reported in this paper. In the one-step process, ferrocyanide salt and excess lime are added ahead of the filter, and the resulting filter cake solidifies by a pozzolanic reaction. The two-step process involves addition of solidifying agents subsequent to filtration. It was found that high surface area diatomaceous synthetic calcium silicate powders, sold commercially as functional fillers and carriers, adsorb nickel isotopes from solution at neutral and slightly basic pH. Addition of the silicates to cooling water allowed removal of the tested metal isotopes (nickel, iron, manganese, cobalt, and cesium) simultaneously at neutral to slightly basic pH. Lime to diatomite ratio was the most influential characteristic of composition on final strength tested, with higher lime ratios giving higher strength. Diatomaceous earth filter aids manufactured without sodium fluxes exhibited higher pozzolanic activity. Pozzolanic filter cake solidified with sodium silicate and a ratio of 0.45 parts lime to 1 part diatomite had compressive strength ranging from 470 to 595 psi at a 90% confidence level. Leachability indices of all tested metals in the solidified waste were acceptable. In light of the typical requirement of removing iron and desirability of control over process pH, a two-step process involving addition of Portland cement to the filter cake may be most generally applicable.

  2. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  3. Processing zirconium alloy used in light water reactors for specified creep rate

    SciTech Connect

    Foster, J.P.; Worcester, S.A.; Comstock, R.J.

    1992-06-30

    This patent describes a method for processing a zirconium alloy, material of a type wherein the material is subjected to a post extrusion meal, a series of an intermediate area reductions and intermediate recrystallization anneals possibly being a late stage beta, a final area reduction, and a final anneal. It comprises controlling creep rate in an alloy consisting essentially of by weight percent, 0.5-2.0 niobium, 0.7-1.5 tin, 0.7-0.28 of at least one of iron, nickel and chromium and up to 20 ppm carbon, and the balance essentially zirconium by subjecting the material to an alpha post extrusion anneal and a final stress relief anneal; subjecting the material to an alpha post extrusion anneal and a final stress relief anneal; subjecting the material to an alpha post extrusion anneal and a final at least partial recrystallization anneal; subjecting the material to an alpha post extrusion anneal and a final at least partial recrystallization anneal; subjecting the material to a beta post extrusion anneal and a final at least partial recrystallization anneal; subjecting the material to a beta post extrusion anneal and a final stress relief anneal; subjecting the material to a beta post extrusion anneal and a final at least partial recrystallization anneal.

  4. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  5. Skin cooling aids cerebrovascular function more effectively under severe than moderate heat stress.

    PubMed

    Lucas, Rebekah A I; Ainslie, Philip N; Fan, Jui-Lin; Wilson, Luke C; Thomas, Kate N; Cotter, James D

    2010-05-01

    Skin surface cooling has been shown to improve orthostatic tolerance; however, the influence of severe heat stress on cardiovascular and cerebrovascular responses to skin cooling remains unknown. Nine healthy males, resting supine in a water-perfusion suit, were heated to +1.0 and +2.0 degrees C elevation in body core temperature (T (c)). Blood flow velocity in the middle cerebral artery (transcranial Doppler ultrasound), mean arterial pressure (MAP; photoplethysmography), stroke volume (SV; Modelflow), total peripheral resistance (TPR; Modelflow), heart rate (HR; ECG) and the partial pressure of end-tidal carbon dioxide (P(ET)CO(2)) were measured continuously during 1-min baseline and 3-min lower body negative pressure (LBNP, -15 mm Hg) when heated without and again with skin surface cooling. Nine participants tolerated +1 degrees C and six participants reached +2 degrees C. Skin cooling elevated (P = 0.004) MAP ~4% during baseline and LBNP at +1 degrees C T (c). During LBNP, skin cooling increased SV (9%; P = 0.010) and TPR (0.9 mm Hg L(-1) min, P = 0.013) and lowered HR (13 b min(-1), P = 0.012) at +1 degrees C T (c) and +2 degrees C T (c) collectively. At +2 degrees C T (c), skin cooling elevated P(ET)CO(2) ~4.3 mm Hg (P = 0.011) and therefore reduced cerebral vascular resistance ~0.1 mm Hg cm(-1) s at baseline and LBNP (P = 0.012). In conclusion, skin cooling under severe heating and mild orthostatic stress maintained cerebral blood flow more effectively than it did under moderate heating, in conjunction with elevated carbon dioxide pressure, SV and arterial resistance. PMID:19946700

  6. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  7. Cold neutron moderator on an upgraded IBR-2 reactor: The first set of results

    NASA Astrophysics Data System (ADS)

    Anan'ev, V. D.; Belyakov, A. A.; Bulavin, M. V.; Verkhoglyadov, A. E.; Kulikov, S. A.; Mukhin, K. A.; Shabalin, E. P.

    2014-02-01

    The first criticality of a new KZ-202 neutron moderator on the IBR-2M reactor is achieved. The moderator consists of thermal and cold units. The former is a room-temperature comb water moderator; the latter, a moderator using a mixture of aromatic hydrocarbons (mesitylene and m-xylene). The cold moderator is filled with granules of this mixture, which are supplied by a cold helium flow, and operates at 30 K. The combination of two units in one moderator makes it possible to simultaneously take the thermal and cold neutron spectra for extracted-beam spectrometers. The arrangement of the thermal and cold moderators is numerically optimized by the Monte Carlo method. The use of the cold moderator allows a 13-fold increase in the cold neutron intensity from its surface.

  8. Electrolytic hydriding of LaFe(13-x)Si(x) alloys for energy efficient magnetic cooling.

    PubMed

    Lyubina, Julia; Hannemann, Ullrich; Ryan, Mary P; Cohen, Lesley F

    2012-04-17

    An effective, low-temperature and readily available electrochemical method for tuning the operation temperature of LaFe(13-x)Si(x)-type alloys is demonstrated. Electrolytically hydrided materials have the same high level magnetic properties as in high temperature gas-phase processed materials and offer an advantage of higher hydrogen absorption rate in the ferromagnetic state. PMID:22431476

  9. Final report on the safety assessment of aluminum silicate, calcium silicate, magnesium aluminum silicate, magnesium silicate, magnesium trisilicate, sodium magnesium silicate, zirconium silicate, attapulgite, bentonite, Fuller's earth, hectorite, kaolin, lithium magnesium silicate, lithium magnesium sodium silicate, montmorillonite, pyrophyllite, and zeolite.

    PubMed

    Elmore, Amy R

    2003-01-01

    irritation study. Magnesium Aluminum Silicate and Sodium Magnesium Silicate caused minimal eye irritation in a Draize eye irritation test. Bentonite caused severe iritis after injection into the anterior chamber of the eyes of rabbits and when injected intralamellarly, widespread corneal infiltrates and retrocorneal membranes were recorded. In a primary eye irritation study in rabbits, Hectorite was moderately irritating without washing and practically nonirritating to the eye with a washout. Rats tolerated a single dose of Zeolite A without any adverse reaction in the eye. Calcium Silicate had no discernible effect on nidation or on maternal or fetal survival in rabbits. Magnesium Aluminum Silicate had neither a teratogenic nor adverse effects on the mouse fetus. Female rats receiving a 20% Kaolin diet exhibited maternal anemia but no significant reduction in birth weight of the pups was recorded. Type A Zeolite produced no adverse effects on the dam, embryo, or fetus in either rats or rabbits at any dose level. Clinoptilolite had no effect on female rat reproductive performance. These ingredients were not genotoxic in the Ames bacterial test system. In primary hepatocyte cultures, the addition of Attapulgite had no significant unscheduled DNA synthesis. Attapulgite did cause significant increases in unscheduled DNA synthesis in rat pleural mesothelial cells, but no significant increase in sister chromosome exchanges were seen. Zeolite particles (<10 microm) produced statistically significant increase in the percentage of aberrant metaphases in human peripheral blood lymphocytes and cells collected by peritoneal lavage from exposed mice. Topical application of Magnesium Aluminum Silicate to human skin daily for 1 week produced no adverse effects. Occupational exposure to mineral dusts has been studied extensively. Fibrosis and pneumoconiosis have been documented in workers involved in the mining and processing of Aluminum Silicate, Calcium Silicate, Zirconium Silicate

  10. CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

    SciTech Connect

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01

    In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

  11. Summary of HTGR (high-temperature gas-cooled reactor) benchmark data from the high temperature lattice test reactor

    SciTech Connect

    Newman, D.F.

    1989-10-01

    The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000{degree}C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of the measurement of k{sub {infinity}} using the unpoisoned technique (developed by R.E. Heineman of PNL) was subjected to extensive peer review within PNL and the General Atomic Company. A number of experiments were conducted at PNL in the Physical Constants Testing Reactor (PCTR) using both the unpoisoned technique and the well-established null reactivity technique that substantiated the equivalence of the measurements by direct comparison. Records of all data from fuel fabrication, the reactor experiments, and the analytical results were compiled and maintained to meet applicable quality assurance standards in place at PNL. Sensitivity of comparisons between measured and calculated k{sub {infinity}}(T) data for various HTGR lattices to changes in neutron cross section data, graphite scattering kernel models, and fuel block loading variations, were analyzed by PNL for the Electric Power Research Institute. As a part of this effort, the fuel rod composition in the dilute {sup 233}UO{sub 2}-ThO{sub 2} HTGR central cell (HTLTR Lattice {number sign}3) was sampled and analyzed by mass spectrometry. Values of k{sub {infinity}} calculated for that lattice were about 5% higher than those measured. Trace quantities of sodium chloride were found in the fuel rod that were equivalent to 22 atom parts-per-million of natural boron.

  12. Experimental characterization of Zircaloy-4 sheet deformation during combined hydriding and corrosion

    NASA Astrophysics Data System (ADS)

    Krebs, B.; Desquines, J.; Busser, V.; Drouan, D.; Zanellato, O.

    2013-04-01

    In the frame of its research activity on nuclear fuel safety, the French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN) has been investigating the dimensional stability of fuel assemblies in Pressurized Water Reactors (PWR). Under normal operating conditions, the dimensional stability is an important performance issue for light water reactor fuel assemblies. The dimensional change results from a combined influence of oxidation, hydriding, irradiation growth and irradiation creep. In the present work, detailed studies carried out on zirconium alloys address the combined influence of oxide layer growth, hydride precipitation and thermal creep. In order to quantify the contribution of each parameter, three kinds of experiments were performed on samples cut in a recrystallized Zircaloy-4 sheet: hydrogen gaseous charging, air oxidation and steam oxidation. These experiments confirm the influence of hydrogen on sample deformation during oxidation. A basic understanding of the key mechanisms affecting the sample strain is proposed.

  13. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  14. SEPARATING LIQUID MODERATOR FROM A SLURRY TYPE REACTOR

    DOEpatents

    Vernon, H.C.

    1961-07-01

    A system for evaporating moderator such as D/sub 2/O from an irradiated slurry or sloution characterized by two successive evaproators is described. In the first of these the most troublesome radioactivity dissipates before the slurry becomes too thick to be pumped out; in the second the slurry, now easier to handle, can be safely reduced to a sludge.

  15. Current status of advanced pelletized cold moderators development for IBR-2M research reactor

    NASA Astrophysics Data System (ADS)

    Kulikov, S.; Belyakov, A.; Bulavin, M.; Mukhin, K.; Shabalin, E.; Verhoglyadov, A.

    2013-03-01

    The world's first advanced pelletized cold neutron moderator is prepared to be put into operation at the IBR-2M pulsed research reactor. It provides long-wavelength neutrons to the most of neutron spectrometers at the beams of the IBR-2M reactor. Aromatic hydrocarbons are used as a material for cold moderators. It is a very attractive material because of its high radiation resistance, good moderating properties, incombustibility, etc. It is shown that the idea of beads transport by a helium flow at cryogenic temperatures is successful. The recent progress and plans for moderator development at the IBR-2M reactor as well as the experimental results of beads transport are discussed in the paper.

  16. Measurement of cryogenic moderator temperature effects in a small heterogeneous thermal reactor

    SciTech Connect

    Hoovler, G.S.; Ball, R.M.; Lewis, R.H.

    1994-12-31

    Past papers have described a critical experiment (CX) built at Sandia National Laboratories to investigate the neutronic behavior of the particle-bed reactor (PBK). Among the experiments previously reported were tests to measure the reactivity effect of uniform temperature variations between 20 and 80{degree}C. This paper describes additional experiments designed to examine the effects of cryogenic moderator temperatures on core reactivity and neutron spectrum. The general importance of temperature effects to the design of the PBR have been previously discussed. A unique feature of the PBR is that the moderator may be at cryogenic temperatures during reactor startup. Because temperature effects in small, heterogeneous thermal reactors can be significant and because we found no integral measurements with cryogenic moderators in such systems, an experiment with a cryogenic moderator was designed and performed in the CX as an extension to the isothermal measurements previously reported.

  17. Documentation of Hybrid Hydride Model for Incorporation into Moose-Bison and Validation Strategy.

    SciTech Connect

    Weck, Philippe F; Tikare, Veena; Schultz, Peter Andrew; Clark, B; Mitchell, J; Glazoff, Michael V.; Homer, Eric R.

    2014-10-01

    This report documents the development, demonstration and validation of a mesoscale, microstructural evolution model for simulation of zirconium hydride δ-ZrH{sub 1.5} precipitation in the cladding of used nuclear fuels that may occur during long-term dry storage. While the Zr-based claddings are manufactured free of any hydrogen, they absorb hydrogen during service, in the reactor by a process commonly termed ‘hydrogen pick-up’. The precipitation and growth of zirconium hydrides during dry storage is one of the most likely fuel rod integrity failure mechanisms either by embrittlement or delayed hydride cracking of the cladding (Hanson et al., 2011). While the phenomenon is well documented and identified as a potential key failure mechanism during long-term dry storage (Birk et al., 2012 and NUREG/CR-7116), the ability to actually predict the formation of hydrides is poor. The model being documented in this work is a computational capability for the prediction of hydride formation in different claddings of used nuclear fuels. This work supports the Used Fuel Disposition Research and Development Campaign in assessing the structural engineering performance of the cladding during and after long-term dry storage. In this work, a model to numerically simulate hydride precipitation at the microstructural scale, in a wide variety of Zr-based claddings, under dry-storage conditions is being developed. It will be used to aid in the evaluation of the mechanical integrity of used fuel rods during dry storage and transportation by providing the structural conditions from the microstructural scale to the continuum scale to engineering component scale models to predict if the used fuel rods will perform without failure under normal and off-normal conditions. The microstructure, especially, the hydride structure is thought to be a primary determinant of cladding failure, thus this component of UFD’s storage and transportation analysis program is critical. The model

  18. A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

    SciTech Connect

    Poston, David I.

    2005-02-06

    The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas--technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the 'customer' - i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

  19. Nuclear data uncertainty propagation for neutronic key parameters of CEA's SFR V2B and CFV sodium fast reactor designs

    SciTech Connect

    Archier, P.; Buiron, L.; De Saint Jean, C.; Dos Santos, N.

    2012-07-01

    This paper presents a nuclear data uncertainty propagation analysis for two CEA's Sodium-cooled Fast Reactor designs: the SFR V2B and CFV cores. The nuclear data covariance matrices are provided by the DER/SPRC/LEPh's nuclear data team (see companion paper) for several major isotopes. From the current status of this analysis, improvements on certain nuclear data reactions are highlighted as well as the need for new specific integral experiments in order to meet the technological breakthroughs proposed by the CFV core. (authors)

  20. Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

    NASA Astrophysics Data System (ADS)

    Dris, Zakaria bin; Mohamed, Abdul Aziz bin; Hamid, Nasri A.; Azman, Azraf; Ahmad, Megat Harun Al Rashid Megat; Jamro, Rafhayudi; Yazid, Hafizal

    2016-01-01

    A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

  1. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    SciTech Connect

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  2. Uranium thorium hydride nuclear fuel

    SciTech Connect

    Simnad, M.T.

    1985-01-15

    A nuclear fuel includes uranium dispersed within a thorium hydride matrix. The uranium may be in the form of particles including fissile and non-fissile isotopes. Various hydrogen to thorium ratios may be included in the matrix. The matrix with the fissile dispersion may be used as a complete fuel for a metal hydride reactor or may be combined with other fuels.

  3. Control system of pelletized cold neutron moderator at the IBR-2 reactor

    NASA Astrophysics Data System (ADS)

    Belyakov, A.; Bulavin, M.; Chernikov, A.; Churakov, A.; Kulikov, S.; Litvinenko, E.; Mukhin, K.; Petrenko, A.; Petukhova, T.; Sirotin, A.; Shabalin, E.; Shirokov, V.; Verhoglyadov, A.

    2015-11-01

    The unique pelletized cold neutron moderator CM-202 at the IBR-2 reactor was put into test operation and have already worked more than 2000 hours. Normal, fast and trouble-free operation of the cryogenic moderator requires strict adherence to technological conditions (fast charging and discharging of moderator chamber, maintenance of required temperature and pressure at different parts of cryogenic system). The system of control and measuring equipment, designed for cryogenic moderator of the IBR-2 reactor, satisfies all the requirements and is simple to use. Access to the system of measuring instruments is organized via network. The working cycles of moderator confirmed the reliability and stable operation of the whole control system.

  4. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    SciTech Connect

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

  5. Research of a Supercritical Pressure Water Cooled Reactor in Korea

    SciTech Connect

    Bae, Yoon-Yeong; Joo, Hyung-Kook; Jang, Jinsung; Jeong, Yong-Hwan; Song, Jin-ho; Yoon, Han-Young; Yoo, Jung-Yul

    2004-07-01

    In this paper the activities on the supercritical pressure water-cooled reactor (SCWR) in Korea are briefly introduced. Four projects on a SCWR are being conducted in Korea. Three of them are supported by the I-NERI program while one is by KAERI. Two of the I-NERI-supported projects concern suitable materials for supercritical pressure and temperature, and radiation environment. The other I-NERI-supported project surveys numerically and experimentally the proper turbulence modeling for the numerical calculation of heat transfer phenomena at a supercritical condition. Heat transfer at a supercritical condition is being studied at KAERI experimentally using carbon dioxide as a coolant. The test loop is to be completed by the end of 2004. (authors)

  6. Scaling analysis for the direct reactor auxiliary cooling system for FHRs

    SciTech Connect

    Lv, Q.; Kim, I. H.; Sun, X.; Christensen, R. N.; Blue, T. E.; Yoder, G.; Wilson, D.; Sabharwall, P.

    2015-04-01

    The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility.

  7. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    SciTech Connect

    Gautier, G. M.; Morin, F.; Dechelette, F.; Sanseigne, E.; Chabert, C.

    2012-07-01

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit

  8. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  9. Pre-orbital criticality safety for the NEPSTP mission

    SciTech Connect

    Sapir, J.; Pelowitz, D.; Streetman, J.R.; Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Lobynstev, V.A.

    1993-11-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) proposed investigating whether launching a Russian Topaz-II space nuclear power system could be done safely and within budget constraints. Functional safety requirements developed for the US Topaz mission mandated that the reactor remain subcritical when immersed in water. Topaz-II is an epithermal, enriched-uranium-fueled, NaK- (liquid metal alloy with 22% sodium and 78% potassium) cooled, and zirconium hydride-moderated reactor. A radial beryllium reflector containing 12 rotatable control drums surrounds the core. The authors prepared a computer model of the Topaz reactor that explicitly represented all major reactor components. Initial analyses indicated that in several water-immersion scenarios, the reactor would not remain subcritical. After additional calculations, modifications were proposed that would assure subcriticality under such conditions. This paper describes the analyses and the proposed modifications.

  10. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

  11. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    SciTech Connect

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

  12. San Diego G and E shows how to make sodium hypochlorite for cooling-water treatment

    SciTech Connect

    Furgerson, S.

    1983-10-01

    There's a trend in power plant cooling water treatment away from gaseous chlorine toward use of a dilute solution of sodium hypochlorite. Some hypochlorite users are finding that savings produced by generating the material on site from seawater instead of purchasing it can pay back the capital costs in two to three years. One of the first plants in the US to generate hypochlorite on site was San Diego Gas and Electric's Encina station. 1 figure.

  13. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOEpatents

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  14. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  15. METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

    DOEpatents

    Vernon, H.C.

    1962-08-14

    A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

  16. Process for purifying zirconium sponge

    SciTech Connect

    Abodishish, H.A.M.; Kimball, L.S.

    1992-03-31

    This patent describes a Kroll reduction process wherein a zirconium sponge contaminated with unreacted magnesium and by-product magnesium chloride is produced as a regulus, a process for purifying the zirconium sponge. It comprises: distilling magnesium and magnesium chloride from: a regulus containing a zirconium sponge and magnesium and magnesium chloride at a temperature above about 800{degrees} C and at an absolute pressure less than about 10 mmHg in a distillation vessel to purify the zirconium sponge; condensing the magnesium and the magnesium chloride distilled from the zirconium sponge in a condenser; and then backfilling the vessel containing the zirconium sponge and the condenser containing the magnesium and the magnesium chloride with a gas; recirculating the gas between the vessel and the condenser to cool the zirconium sponge from above about 800{degrees} C to below about 300{degrees} C; and cooling the recirculating gas in the condenser containing the condensed magnesium and the condensed magnesium chloride as the gas cools the zirconium sponge to below about 300{degrees} C.

  17. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    SciTech Connect

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-07-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material {sup 233}U which shows superior characteristics in the thermal region. The Shippingport reactor used {sup 233}U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, {eta} value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required {sup 233}U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of {sup 233}U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  18. CFD analyses of natural circulation in the air-cooled reactor cavity cooling system

    SciTech Connect

    Hu, R.; Pointer, W. D.

    2013-07-01

    The Natural Convection Shutdown Heat Removal Test Facility (NSTF) is currently being built at Argonne National Laboratory, to evaluate the feasibility of the passive Reactor Cavity Cooling System (RCCS) for Next Generation Nuclear Plant (NGNP). CFD simulations have been applied to evaluate the NSTF and NGNP RCCS designs. However, previous simulations found that convergence was very difficult to achieve in simulating the complex natural circulation. To resolve the convergence issue and increase the confidence of the CFD simulation results, additional CFD simulations were conducted using a more detailed mesh and a different solution scheme. It is found that, with the use of coupled flow and coupled energy models, the convergence can be greatly improved. Furthermore, the effects of convection in the cavity and the effects of the uncertainty in solid surface emissivity are also investigated. (authors)

  19. Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1973-01-01

    An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

  20. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    SciTech Connect

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  1. REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

    DOEpatents

    Rodin, M.B.; Carter, J.C.

    1962-05-15

    A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

  2. 158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  3. XRD and NMR investigation of Ti-compound formation in solution-doping of sodium aluminum hydrides: Solubility of Ti in NaAlH4 crystals grown in THF

    SciTech Connect

    Majzoub, E H; Herberg, J L; Stumpf, R; Spangler, S; Maxwell, R S

    2004-08-26

    Sodium aluminum hydrides have gained attention due to their high hydrogen weight percent (5.5% ideal) compared to interstitial hydrides, and as a model for hydrides with even higher hydrogen weight fraction. The purpose of this paper is to investigate the Ti-compounds that are formed under solution-doping techniques, such as wet doping in solvents such as tetrahydrofuran (THF). Compound formation in Ti-doped sodium aluminum hydrides is investigated using x-ray diffraction (XRD) and magic angle spinning (MAS) nuclear magnetic resonance (NMR). We present lattice parameter measurements of crushed single crystals, which were exposed to Ti during growth. Rietveld refinements indicate no lattice parameter change and thus no solubility for Ti in NaAlH{sub 4} by this method of exposure. In addition, x-ray diffraction data indicate that no Ti substitutes in NaH, the final decomposition product for the alanate. Reaction products of completely reacted (33.3 at. %-doped) samples that were solvent-mixed or mechanically milled are investigated. Formation of TiAl{sub 3} is observed in mechanically milled materials, but not solution mixed samples, where bonding to THF likely stabilizes Ti-based nano-clusters. The Ti in these clusters is activated by mechanical milling.

  4. R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

    SciTech Connect

    Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

    2011-07-01

    In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

  5. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

    SciTech Connect

    Welty, R. K.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis.

  6. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  7. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  8. High temperature gas-cooled reactor: gas turbine application study

    SciTech Connect

    Not Available

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  9. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  10. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  11. Measurement of cold neutron spectra using a model cryogenic moderator of the IBR-2M reactor

    NASA Astrophysics Data System (ADS)

    Kulikov, S. A.; Kalinin, I. V.; Morozov, V. M.; Novikov, A. G.; Puchkov, A. V.; Chernikov, A. N.; Shabalin, E. P.

    2010-01-01

    The method and results of an experiment to determine the cold neutron spectrum from solid mesitylene at moderator temperatures of 10-50 K are presented. This study was performed at the DIN-2PI spectrometer of the IBR-2 reactor. The objective of the study was to verify the system of constants used in the Monte Carlo simulation of cryogenic neutron moderators of the IBR-2M reactor and to obtain the cold neutron yield as a function of the moderator temperature. Satisfactory agreement between the experimental and calculated neutron spectra at a mesitylene temperature of 20 K has been obtained; the ratio of cold neutron intensities at 10 and 50 K is ˜1.8.

  12. Development of new reactor fuel materials: hydrogenation properties of UThZr alloys and neutron irradiation effects on their hydrides

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takuya; Suwamo, Hadi; Kayano, Hideo; Yamawaki, Michio

    1997-08-01

    Hydrogen absorption properties of four UThZr alloys with compositions of 2:1:6, 1:1:4, 1:2:6 and 1:4:10 in U:Th:Zr ratio were examined for developing a new UTh mixed hydride fuel. It was revealed by hydrogen absorption measurement that for all the specimens hydrogen capacities normalized by the sum of Th and Zr contents were similar to or higher than that of the UZrH 2 - x alloy, TRIGA fuel, at temperatures from 773 to 1073 K and under hydrogen pressures from 10 2 to 10 5 Pa. Regarding the microstructure, the alloy hydrides consisted of three phases; α-U, ZrH 2 - x and ThZr 2H 7 - x , which are finely and homogeneously mixed with each other probably because they were formed from one solid solution phase stable at high temperatures. In the case of Th-rich alloys the α-U was about 1 μm in diameter and dispersed in the bulk of ThZr 2 H 7 - x and ZrH 2 - x. Such microstructure is quite similar to that of UZrH 2 - x fuel. The hydrides irradiated to 7.4 × 10 23 n/m 2 in the reactor showed a high phase stability.

  13. Characterization of the sodium void reactivity effect for advanced liquid metal reactor fuels

    SciTech Connect

    Kessler, S.F.

    1993-12-01

    This report discusses the problems of a large positive sodium void reactivity effect in liquid metal reactors which have received increased attention following the accident at Chernobyl, a light water reactor with a positive coolant void coefficient. While the probability of voiding sodium is small, a large positive sodium void reactivity effect is, in many minds, unacceptable. Analyses were performed on models of an advanced liquid metal reactors to determine the effects fuel type have on the sodium void reactivity effect. Three fuel types were considered; metal, oxide, and nitride. Calculations were performed using three-dimensional, multigroup diffusion theory. Two programs were developed to aid the analyses. One calculated the capture-to-fission ratio and the other calculated reaction rates of selected materials. A one-group equation was derived to determine a theoretical basis for the sodium void reactivity effect. An option was presented for a shortened core having a near-zero sodium-void worth. The effect on the sodium void reactivity effect of using actinides as fuel is also considered.

  14. Characteristics and Applications of Metal Hydrides

    NASA Technical Reports Server (NTRS)

    Egan, G. J.; Lynch, F. E.

    1987-01-01

    Report discusses engineering principles of uses of metal hydrides in spacecraft. Metal hydrides absorb, store, pump, compress, and expand hydrogen gas. Additionally, they release or absorb sizeable amounts of heat as they form and decompose - property adapted for thermal-energy management or for propulsion. Describes efforts to: Identify heat sources and sinks suitable for driving metal hydride thermal cycles in spacecraft; develop concepts for hydride subsystems employing available heating and cooling methods; and produce data base on estimated sizes, masses, and performances of hydride devices for spacecraft.

  15. Simplified thermochemistry of oxygen in lithium and sodium for liquid metal cooling systems

    NASA Technical Reports Server (NTRS)

    Tower, L. K.

    1972-01-01

    Plots of oxygen chemical potential against composition of lithium-oxygen solutions and sodium-oxygen solutions for a range of temperature were constructed. For each liquid metal two such plots were prepared. For one plot ideal solution behavior was assumed. For the other plot, existing solubility limit data for oxygen in the liquid metal were used to determine a first-order term for departure from ideality. The use of the plots in evaluating the oxygen gettering capability of refractory metals in liquid metal cooling systems is illustrated by a simple example involving lithium, oxygen, and hafnium.

  16. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  17. Low potassium enhances sodium uptake in red-beet under moderate saline conditions

    NASA Technical Reports Server (NTRS)

    Subbarao, G. V.; Wheeler, R. M.; Stutte, G. W.; Levine, L. H.; Sager, J. C. (Principal Investigator)

    2000-01-01

    Due to the discrepancy in metabolic sodium (Na) requirements between plants and animals, cycling of Na between humans and plants is limited and critical to the proper functioning of bio-regenerative life support systems, being considered for long-term human habitats in space (e.g., Martian bases). This study was conducted to determine the effects of limited potassium (K) on growth, Na uptake, photosynthesis, ionic partitioning, and water relations of red-beet (Beta vulgaris L. ssp. vulgaris) under moderate Na-saline conditions. Two cultivars, Klein Bol, and Ruby Queen were grown for 42 days in a growth chamber using a re-circulating nutrient film technique where the supplied K levels were 5.0, 1.25, 0.25, and 0.10 mM in a modified half-strength Hoagland solution salinized with 50 mM NaCl. Reducing K levels from 5.0 to 0.10 mM quadrupled the Na uptake, and lamina Na levels reached -20 g kg-1 dwt. Lamina K levels decreased from -60 g kg-1 dwt at 5.0 mM K to -4.0 g kg-1 dwt at 0.10 mM K. Ruby Queen and Klein Bol responded differently to these changes in Na and K status. Klein Bol showed a linear decline in dry matter production with a decrease in available K, whereas for cv. Ruby Queen, growth was stimulated at 1.25 mM K and relatively insensitive to a further decreases of K down to 0.10 mM. Leaf glycinebetaine levels showed no significant response to the changing K treatments. Leaf relative water content and osmotic potential were significantly higher for both cultivars at low-K treatments. Leaf chlorophyll levels were significantly decreased at low-K treatments, but leaf photosynthetic rates showed no significant difference. No substantial changes were observed in the total cation concentration of plant tissues despite major shifts in the relative Na and K uptake at various K levels. Sodium accounted for 90% of the total cation uptake at the low K levels, and thus Na was likely replacing K in osmotic functions without negatively affecting the plant water status, or

  18. Research Reactor Benchmarks

    SciTech Connect

    Ravnik, Matjaz; Jeraj, Robert

    2003-09-15

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given.

  19. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T.

    2012-07-01

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

  20. Cold source moderator vessel development for the High Flux Isotope Reactor: Thermal-hydraulic studies

    SciTech Connect

    Williams, P.T.; Lucas, A.T.; Wendel, M.W.

    1998-07-01

    A project is underway at Oak Ridge National Laboratory (ORNL) to design, test, and install a cold neutron source facility in the High Flux Isotope Reactor (HFIR). This new cold source employs supercritical hydrogen at cryogenic temperatures both as the medium for neutron moderation and as the working fluid for removal of internally-generated nuclear heating. The competing design goals of minimizing moderator vessel mass and providing adequate structural integrity for the vessel motivated the requirement of detailed multidimensional thermal-hydraulic analyses of the moderator vessel as a critical design subtask. This paper provides a summary review of the HFIR cold source moderator vessel design and a description of the thermal-hydraulic studies that were carried out to support the vessel development.

  1. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  2. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    SciTech Connect

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  3. Trapping and cooling of sodium atoms for assembly of dipolar molecules

    NASA Astrophysics Data System (ADS)

    Yu, Yichao; Hutzler, Nicholas R.; Liu, Lee R.; Zhang, Jessie T.; Ni, Kang-Kuen

    2016-05-01

    In order to create a diatomic molecule with a large electric dipole moment, it is generally necessary to use atoms with very different electronegativities. In the context of bi-alkali molecules, this means combining a light alkali atom with a heavy one. This is the reason why we use NaCs in our molecule assembler experiment; NaCs has the largest induced dipole moment in few kV/cm lab fields. However, the use of sodium atoms also poses challenges. The higher Doppler temperature and lack of efficient D2 polarization gradient cooling increases the necessary depth of our optical dipole (tweezer) traps. The lack of a convenient magic wavelength for the dipole trap creates a large AC stark shift on the optical transition as well as additional heating mechanisms. The light mass of the sodium, and therefore larger Lamb-Dicke parameter and higher recoil temperature, makes it more difficult to perform efficient Raman sideband cooling on the atom in the trap. I will discuss the techniques we use to overcome these challenges, in particular a method to eliminate the light shifts and associated heating mechanisms in tight optical traps.

  4. Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

    NASA Astrophysics Data System (ADS)

    Bays, Samuel Eugene

    2008-10-01

    In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this

  5. Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

    DOEpatents

    Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

    1977-01-01

    A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

  6. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  7. Ballooning of CANDU pressure tubes: Hydride blisters and iodine

    SciTech Connect

    Forrest, C.F.

    1997-12-31

    Reports on a project that continues a test program to determine the effects of in-service degradation and abnormal environmental conditions on the ballooning behavior of pressure tubes in CANDU nuclear reactors. Investigators tested nine zirconium-niobium pressure tube specimens to study the effects of hydride blisters and uniformly distributed hydrogen, and the composition of the internal pressurizing gas, including argon, steam, a steam-iodine mixture, and hydrogen. Specimen internal pressure and temperature ramp rate ranged from one megapascal and one degree K per second to 9.6 megapascals and 35 K per second. Three reference tests were carried out with as-received material. Temperatures on the outside and inside surfaces of the specimens, and circumferential and longitudinal strains, were recorded during the transients. Post-test longitudinal, circumferential, and wall thickness strains were also measured.

  8. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    DOEpatents

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  9. Low density metal hydride foams

    DOEpatents

    Maienschein, Jon L.; Barry, Patrick E.

    1991-01-01

    Disclosed is a low density foam having a porosity of from 0 to 98% and a density less than about 0.67 gm/cc, prepared by heating a mixture of powered lithium hydride and beryllium hydride in an inert atmosphere at a temperature ranging from about 455 to about 490 K for a period of time sufficient to cause foaming of said mixture, and cooling the foam thus produced. Also disclosed is the process of making the foam.

  10. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect

    1990-05-01

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  11. METHOD FOR REMOVING SODIUM OXIDE FROM LIQUID SODIUM

    DOEpatents

    Bruggeman, W.H.; Voorhees, B.G.

    1957-12-01

    A method is described for removing sodium oxide from a fluent stream of liquid sodium by coldtrapping the sodium oxide. Apparatus utilizing this method is disclosed in United States Patent No. 2,745,552. Sodium will remain in a molten state at temperatures below that at which sodium oxide will crystallize out and form solid deposits, therefore, the contaminated stream of sodium is cooled to a temperature at which the solubility of sodium oxide in sodium is substantially decreased. Thereafter the stream of sodium is passed through a bed of stainless steel wool maintained at a temperature below that of the stream. The stream is kept in contact with the wool until the sodium oxide is removed by crystal growth on the wool, then the stream is reheated and returned to the system. This method is useful in purifying reactor coolants where the sodium oxide would otherwise deposit out on the walls and eventually plug the coolant tubes.

  12. Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Breitwieser, R.; Niederauer, G. F.

    1974-01-01

    There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

  13. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Not Available

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant's structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  14. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Donna Post Guillen

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  15. Generic component failure data base for light water and liquid sodium reactor PRAs (probabilistic risk assessments)

    SciTech Connect

    Eide, S.A.; Chmielewski, S.V.; Swantz, T.D.

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs). The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates. Using this approach, most of the failure rates are based on actual plant data rather than existing estimates. 21 refs., 9 tabs.

  16. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    DOEpatents

    Hutter, Ernest; Pardini, John A.

    1977-03-15

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads.

  17. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect

    Greenspan, E

    2006-04-30

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the

  18. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    DOEpatents

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  19. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Strydom, Gerhard

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  20. Rehydration with drinks differing in sodium concentration and recovery from moderate exercise-induced hypohydration in man.

    PubMed

    Merson, Stuart J; Maughan, Ronald J; Shirreffs, Susan M

    2008-07-01

    To investigate how differing moderate sodium chloride concentrations affect rehydration after exercise and subsequent exercise capacity, eight males lost 1.98 +/- 0.1% body mass exercising in the heat, then consumed one of four drinks in a volume equivalent to 150% of mass loss. Drinks were identical except for sodium chloride content (1 +/- 1, 31 +/- 1, 40 +/- 1, 50 +/- 1 mmol/l). After 4 h recovery subjects cycled for 5 min at 70% VO(2peak) then at 95% VO(2peak) until volitional fatigue. Urine output was inversely related to sodium intake: more was produced with the 1 than the 40 and 50 mmol/l drinks (P < 0.01). Time to exhaustion in the exercise capacity test was not different between treatments (P = 0.883). The addition of 40 or 50 mmol/l of sodium chloride to a rehydration beverage reduced subsequent urine output, thereby providing more effective rehydration than a sodium-free drink. This did not, however, result in improved performance 4 h after the end of the rehydration period. PMID:18463891

  1. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    SciTech Connect

    Sulaiman, S. A. Dominguez-Ontiveros, E. E. Alhashimi, T. Budd, J. L. Matos, M. D. Hassan, Y. A.

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  2. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    NASA Astrophysics Data System (ADS)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  3. Lead-cooled fast reactor use in future equilibrium energy production

    SciTech Connect

    Sekimoto, Hiroshi; Kuznetsov, V.V.

    1994-12-31

    The design of a lead cooled fast reactor is discussed. In previous works, general characteristics of future nuclear equilibrium energy utilization have been investigated where the toxic radioactive materials are confined in a nuclear center. Natural uranium and/or thorium, are supplied to the center as a fuel fed to fission reactors. All of the actinides are recycled in the reactor. The end products of the heavy-isotope decay series (lead and bismuth) and the stable fission products are taken out of the center. The discharged short- and middle-life fission products are cooled until they die out in the center, then the previously described process is used.

  4. Method and apparatus for enhancing reactor air-cooling system performance

    DOEpatents

    Hunsbedt, Anstein

    1996-01-01

    An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

  5. Method and apparatus for enhancing reactor air-cooling system performance

    DOEpatents

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  6. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    SciTech Connect

    DeHart, Mark D; Pritchard, Megan L

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of

  7. Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C

    SciTech Connect

    Ian Mckirdy

    2010-12-01

    This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

  8. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    NASA Astrophysics Data System (ADS)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  9. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L.; Sun, X.

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  10. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    NASA Astrophysics Data System (ADS)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  11. Hydride compositions

    DOEpatents

    Lee, Myung W.

    1995-01-01

    A composition for use in storing hydrogen, and a method for making the composition. The composition comprises a mixture of two or more hydrides, each hydride having a different series of hydrogen sorption isotherms that contribute to the overall isotherms of the mixture. The hydrides are chosen so that the isotherms of the mixture have regions wherein the hydrogen equilibrium pressure increases with increasing hydrogen, preferably linearly. The isotherms of the mixture can be adjusted by selecting hydrides with different isotherms and by varying the amounts of the individual hydrides, or both. Preferably, the mixture is made up of hydrides that have isotherms with substantially flat plateaus and in nearly equimolar amounts. The composition is activated by degassing, exposing to hydrogen and then heating at a temperature below the softening temperature of any of the. constituents so that their chemical and structural integrity is preserved. When the composition is used to store hydrogen, its hydrogen content can be found simply by measuring P.sub.H.sbsb.2 and determining H/M from the isothermic function of the composition.

  12. Hydride compositions

    DOEpatents

    Lee, Myung, W.

    1994-01-01

    Disclosed are a composition for use in storing hydrogen and a method for making the composition. The composition comprises a mixture of two or more hydrides, each hydride having a different series of hydrogen sorption isotherms that contribute to the overall isotherms of the mixture. The hydrides are chosen so that the isotherms of the mixture have regions wherein the H equilibrium pressure increases with increasing hydrogen, preferably linearly. The isotherms of the mixture can be adjusted by selecting hydrides with different isotherms and by varying the amounts of the individual hydrides, or both. Preferably, the mixture is made up of hydrides that have isotherms with substantially flat plateaus and in nearly equimolar amounts. The composition is activated by degassing, exposing to H, and then heating below the softening temperature of any of the constituents. When the composition is used to store hydrogen, its hydrogen content can be found simply by measuring P{sub H}{sub 2} and determining H/M from the isothermic function of the composition.

  13. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N.; Kannan, S.E.

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  14. Simplified analysis of PRISM RVACS (Reactor Vessel Auxiliary Cooling System) performance without liner spill-over

    SciTech Connect

    Van Tuyle, G.J.

    1990-01-01

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs.

  15. Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

    SciTech Connect

    S. D. Herrmann; L. A. Wurth; N. J. Gese

    2013-09-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

  16. Radiation effects on corrosion of zirconium alloys

    SciTech Connect

    Johnson, A.B. Jr.

    1989-06-01

    From the wide use of zirconium alloys as components in nuclear reactors, has come clear evidence that reactor radiation is a major corrosion parameter. The evidence emerges from comparisons of zirconium alloy corrosion behavior in different reactor types, for example, BWRs versus PWRs and in corresponding reactor loop chemistries; also, oxidation rates differ with location along components such as fuel rods and reactor pressure tubes. In most respects, oxidation effects on power reactor components are paralleled by oxidation behavior on specimens exposed to radiation in reactor loops.

  17. Under-sodium viewing technology for improvement of fast-reactor safeguards

    SciTech Connect

    Beddingfield, David H; Gerhart, Jeremy J; Kawakubo, Yoko

    2009-01-01

    The current safeguards approach for fast reactors relies exclusively on maintenance of continuity of knowledge to track the movement of fuel assemblies through these facilities. The remote handling of fuel assemblies, the visual opacity of the liquid metal coolant. and the chemical reactivity of sodium all combine and result in significant limitations on the available options to verify fuel assembly identification numbers or the integrity of these assemblies. These limitations also serve to frustrate attempts to restore the continuity-of-knowledge in instances where the information is under a variety of scenarios. The technology of ultrasonic under-sodium viewing offers new options to the safeguards community for recovering continuity-of-knowledge and applying more traditional item accountancy to fast reactor facilities. We have performed a literature review to investigate the development of under-sodium viewing technologies. In this paper we will summarize our findings and report the state of development of this technology and we will present possible applications to the fast reactor system to improve the existing safeguards approach at these reactors and in future fast reactors.

  18. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    NASA Astrophysics Data System (ADS)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  19. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  20. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  1. Minimization of corrosion using activated sodium bromide in a medium-size cooling tower

    SciTech Connect

    Nalepa, C.J.; Moore, R.M.; Golson, G.L.; Wolfe, T.W.; Puckorius, P.R.

    1996-07-01

    The cooling tower at the Albermarle Process Development Center in Baton Rouge, Louisiana, historically used chlorine as a biocide in combination with phosphorus-based corrosion/scale inhibitors. Although this regimen provided biocontrol, sludge and iron buildup was a problem in low-velocity, small cross-sectional areas of piping. A general cleanup of the system was performed in April 1995. This cleanup was followed with a switch to a two-component corrosion inhibitor/dispersant package. Alternate biocides were evaluated at this time. Activated sodium bromide was found to be particularly effective in this tower, which operates at pH {approximately}8.4. Relative to chlorine, the use of activated sodium bromide led to a decrease in general and pitting corrosion on mild steel. The reduced corrosion appears to be due to a combination of both chemical (less attack on passivated metal surfaces) and biological factors (better control of heterotrophic and sessile bacteria). These conclusions are supported by chemical analyses, corrosion meter and coupon data, dip slides, biological activity reaction tests, and visual observations of the tower sump and heat exchanger surfaces.

  2. Case study: Minimization of corrosion using activated sodium bromide in a medium-size cooling tower

    SciTech Connect

    Nalepa, C.J.; Moore, R.M.; Golson, G.L.; Wolfe, T.W.; Puckorius, P.R.

    1996-10-01

    The process loop cooling tower at the Albemarle Process Development Center in Baton Rouge, LA has historically used chlorine as the biocide together with industry accepted phosphorus-based corrosion/scale inhibitors. Although this regimen provided biocontrol, sludge and iron build-up was a recurring problem, especially in low-velocity, small cross-sectional areas of piping. A general clean-up of the system was performed in April, 1995. This clean-up was followed with a switch to a two-component corrosion inhibitor/dispersant package. It was decided to study alternate biocides as well at this time. Activated sodium bromide was found to be particularly effective in this tower, which operates at pH {approximately}8.4. Relative to chlorine, the use of activated sodium bromide led to a decrease in general and pitting corrosion on mild steel while maintaining prior performance on admiralty brass. The reduced corrosion appears to be due to a combination of both chemical (less attack on passivated metal surfaces) and biological factors (better control of heterotrophic and sessile bacteria). These conclusions are supported by chemical analyses, corrosion meter and coupon data, dip slides, BART (biological activity reaction test) tests, and visual observations of the tower sump and heat exchanger surfaces.

  3. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    SciTech Connect

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  4. Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors

    NASA Astrophysics Data System (ADS)

    Kotlyarevskiy, S. G.; Pavliuk, A. O.; Zakharova, E. V.; Volkova, A. G.

    2016-06-01

    The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products: 93mNb and the long-lived 94Nb, 93Zr radionuclides. Radionuclides of 60Co, 137Cs, 90Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for 60Co and 137Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.

  5. Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

    NASA Technical Reports Server (NTRS)

    Nast, T. C.

    1983-01-01

    The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

  6. THE VALUE OF HELIUM-COOLED REACTOR TECHNOLOGIES OF NUCLEAR WASTE

    SciTech Connect

    C. RODRIGUEZ; A. BAXTER

    2001-03-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations.

  7. Transfer of Amide and 2-Methoxyethoxy Groups and Sodium Encapsulation in the Reaction of TaCl3[N(TMS)2]2 with Sodium Bis(2-methoxyethoxy)aluminum Hydride: X-ray Structure of [NaAl{N(TMS)2}(OCH2CH2OMe)3]2

    SciTech Connect

    Huang, Shih-Huang Huang; Wang, Xiaoping; Richmond, Michael G.

    2009-01-01

    The reaction between the tantalum compound TaCl3[N(TMS)2]2 and the hydridic reducing agent sodium bis(2-methoxyethoxy)aluminum hydride (Vitride) has been investigated in toluene solution at room temperature and found to afford the dimeric aluminate complex [NaAl{N(TMS)2}(OCH2CH2OMe)3]2 as the sole isolable product. The molecular structure of the product establishes the existence of a four-coordinate aluminum atom and the formal transfer of the 2-methoxyethoxy and bis(trimethylsilyl)amide groups to the aluminate product. The aggregation of two NaAl{N(TMS)2}(OCH2CH2OMe)3 units serves to bind the two sodium cations in a crown-ether fashion through six ancillary oxygen atoms.

  8. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1990-01-01

    A landmark safety test has been conducted at the AVR-reactor, a high-temperature gas-cooled reactor (HTGR) in the Federal Republic of Germany owned by the Arbeitsgemeinschaft Versuchsreaktor, AVR in Juelich. The 46-MW(t), 15-MW(e) AVR reactor was subjected to a simulated loss-of-coolant accident (LOCA), a very severe occurrence in which the coolant escapes from the reactor core and no emergency system provides coolant flow to the core. The test, which demonstrated the inherently safe response of this reactor to a LOCA, marked the first time ever that a reactor has been intentionally subjected to loss-of-coolant conditions without emergency cooling. Oak Ridge National Laboratory (ORNL) and General Atomics participated in the test by working with AVR staff by jointly performing the analyses needed to obtain the license to conduct the test and by performing post test analyses. This participation was carried out under the cooperative AVR Subprogram which is conducted within the US/FRG Agreement for Cooperation in Gas-Cooled Reactor Development. 7 figs.

  9. Emergency makeup flow model for the K-reactor cooling water basin

    SciTech Connect

    Barbour, K.L.

    1994-12-31

    The Savannah River site installed the K-reactor cooling tower in 1993 to replace river water supplied to a 25-million-gal cooling basin with cooling tower recirculation. The reactor accident safety analysis assumes that cooling water recirculation is lost during the accident and basin level will drop. Emergency river water supply makeup valves will be opened manually to restore basin makeup and level and maintain shutdown safety. A hydraulic model scopes out valve flow response as the valves are opened. Scoping objectives are (a) valve flow rate response, (b) volumetric makeup with time, and (c) total volumetric makeup effect on basin emergency operating operating procedures. Model results could influence basin emergency operating procedures development before actual field test data are obtained.

  10. Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

    NASA Astrophysics Data System (ADS)

    Varseev, E. V.; Alekseev, V. V.

    2014-11-01

    The mass transfer of the products of corrosion of the steel surface of the sodium loop of a fast nuclear power reactor was investigated for the purpose of optimization of its parameters. The problem of deposition of the corrosion products on the surface of the heat-exchange unit of the indicated loop was considered. Experimental data on the rate of accumulation of deposits in the channel of this unit and results of the dispersion analysis of the suspensions contained in the sodium coolant are presented.

  11. A modular approach to lead-cooled reactors modelling

    NASA Astrophysics Data System (ADS)

    Casamassima, V.; Guagliardi, A.

    2008-06-01

    After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead.

  12. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor

    NASA Astrophysics Data System (ADS)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  13. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  14. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOEpatents

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  15. Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

    SciTech Connect

    Hine, R.E.

    1980-09-01

    This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated.

  16. HTGR (High Temperature Gas-Cooled Reactor) ingress analysis using MINET

    SciTech Connect

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs.

  17. Loss-of-coolant accident experiment at the AVR (Arbeitsgemeinschaft Versuchsreaktor) gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1989-11-01

    Loss of coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into modular gas-cooled reactor designs, loss-of-coolant accident (LOCA) simulation tests were conducted with the 15-MW(electric), 46-MW(thermal), pebble-bed, high-temperature Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany (FRG). This is the only nuclear power plant ever to have been intentionally subjected to LOCa conditions. Oak Ridge National Laboratory participation in the preparation and conduct of the tests was carried out within the U.S./FRG Agreement for Cooperation in Gas-Cooled Reactor Development.

  18. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    DOEpatents

    DeVolpi, A.

    1984-07-20

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  19. Results of NDE Technique Evaluation of Clad Hydrides

    SciTech Connect

    Dennis C. Kunerth

    2014-09-01

    This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used to detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing

  20. Filter/moderator system for a BNCT beam of epithermal neutrons at nuclear reactor MARIA

    NASA Astrophysics Data System (ADS)

    Tyminska, Katarzyna

    2009-01-01

    Boron Neutron Capture Therapy is a very promising form of cancer therapy, consisting in irradiating a stable isotope of boron (10B) concentrated in tumor cells with a low energy neutron beam. This technique makes it possible to destroy tumor cells, leaving healthy tissues practically unaffected. In order to carry out the therapy in the proper way, the proper range of the neutron beam energy has to be chosen. In this paper we present a filter/moderator system modeled with MCNP code in order to obtain an epithermal neutron beam for BNCT post at MARIA reactor in Swierk.

  1. System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

    SciTech Connect

    Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

    2004-03-15

    Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 s after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.

  2. System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

    SciTech Connect

    Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki; Kazuhiro Ohyama

    2002-07-01

    Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heat removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)

  3. Process of precipitating zirconium or hafnium from spent pickling solutions

    SciTech Connect

    Aguilar, C.L.; Walker, R.G.

    1991-12-31

    This patent describes a process for regenerating a spent, fluoride-containing pickle solution used in the pickling of zirconium or hafnium metal or their alloys. It comprises adding to the spent pickle solution a sufficient amount of sodium sulfate to precipitate sodium zirconium or hafnium fluoride.

  4. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  5. Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

    SciTech Connect

    Kroeger, P.G.; Chan, B.C.

    1985-01-01

    Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the CB atmosphere temperature and pressure, as overpressurization is the most likely failure mode.

  6. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Williamson, Joshua; Peters, Curtis D.; Brown, Nicholas; Jablonski, Jennifer

    2005-02-06

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  7. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

    2005-02-01

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  8. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  9. Neutronic Assessment of Transmutation Target Compositions in Heterogeneous Sodium Fast Reactor Geometries

    SciTech Connect

    Samuel E. Bays; Rodolfo M. Ferrer; Michael A. Pope; Benoit Forget; Mehdi Asgari

    2008-02-01

    The sodium fast reactor is under consideration for consuming the transuranic waste in the spent nuclear fuel generated by light water reactors. This work is concerned with specialized target assemblies for an oxide-fueled sodium fast reactor that are designed exclusively for burning the americium and higher mass actinide component of light water reactor spent nuclear fuel (SNF). The associated gamma and neutron radioactivity, as well as thermal heat, associated with decay of these actinides may significantly complicate fuel handling and fabrication of recycled fast reactor fuel. The objective of using targets is to isolate in a smaller number of assemblies these concentrations of higher actinides, thus reducing the volume of fuel having more rigorous handling requirements or a more complicated fabrication process. This is in contrast to homogeneous recycle where all recycled actinides are distributed among all fuel assemblies. Several heterogeneous core geometries were evaluated to determine the fewest target assemblies required to burn these actinides without violating a set of established fuel performance criteria. The DIF3D/REBUS code from Argonne National Laboratory was used to perform the core physics and accompanying fuel cycle calculations in support of this work. Using the REBUS code, each core design was evaluated at the equilibrium cycle condition.

  10. In situ hydrogen loading on zirconium powder

    PubMed Central

    Maimaitiyili, Tuerdi; Blomqvist, Jakob; Steuwer, Axel; Bjerkén, Christina; Zanellato, Olivier; Blackmur, Matthew S.; Andrieux, Jérôme; Ribeiro, Fabienne

    2015-01-01

    For the first time, various hydride phases in a zirconium–hydrogen system have been prepared in a high-energy synchrotron X-ray radiation beamline and their transformation behaviour has been studied in situ. First, the formation and dissolution of hydrides in commercially pure zirconium powder were monitored in real time during hydrogenation and dehydrogenation, then whole pattern crystal structure analysis such as Rietveld and Pawley refinements were performed. All commonly reported low-pressure phases presented in the Zr–H phase diagram are obtained from a single experimental arrangement. PMID:26134803

  11. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  12. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  13. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  14. New approches for high temperature gas cooled reactors (HTGRs)

    SciTech Connect

    Kasten, P.R.; Cleveland, J.C.; Bowers, H.I.

    1984-01-01

    Several approaches are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the power level of the reactor and increase ability to use passive methods for removing afterheat energy following extreme accidents. One approach most fully discussed in this paper is represented by modular HTRs for which the unit size and design are constrained such that extreme accidents do not result in significant release of radioactivity from the reactor circuit. Through such an approach, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant and achieve an economic system. Attaining such performance should provide low investment risk to the owner.

  15. Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure

    SciTech Connect

    Abu Khalid Rivai; Minoru Takahashi

    2006-07-01

    Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with {sup 235}U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without re-shuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC. (authors)

  16. Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

    NASA Astrophysics Data System (ADS)

    Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

    2014-04-01

    The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

  17. Response of biodegradation characteristics of unacclimated activated sludge to moderate pressure in a batch reactor.

    PubMed

    Xu, Rui-Xiao; Li, Bing; Zhang, Yong; Si, Ling; Zhang, Xian-Qiu; Xie, Biao

    2016-04-01

    This study was aimed to investigate the effect of moderate pressure on unacclimated activated sludge. Process of organic degradation, variation of carbon dioxide (CO2) concentration of off-gas and characteristics of extracellular polymeric substances (EPS) of activated sludge were analyzed using pressure-atmospheric comparative experiments in bench-scale batch reactors. It was found that moderate pressure increased the degradation rate more dramatically when the biological process ran under a higher organic load with much more oxygen demand, which illuminated that applications of the pressurized method to high concentration organic wastewaters would be more reasonable and practicable. High oxygen transfer impetus increased utilization of oxygen which not only promoted the biodegradation of organics in wastewater, but also led to more EPS consumption in activated sludge. CO2 concentration of off-gas was lower in the earlier stage due to CO2 being pressed into the liquid phase and converted into inorganic carbon (IC). More CO2 emission was observed during the pressurized aerobic process 160 min later. EPS in pressurized reactor was much lower, which may be an important way of sludge reduction by pressurized technology. PMID:26802261

  18. Shield materials recommended for space power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

  19. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

    NASA Technical Reports Server (NTRS)

    Martin, James J.; Reid, Robert S.

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

  20. New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors

    SciTech Connect

    Greenspan, Ehud; Hejzlar, Pavel; Sekimoto, Hiroshi; Toshinsky, Georgy; Wade, David

    2005-08-15

    Fast reactors cooled by lead or lead-bismuth alloy offer new interesting fuel cycle and fuel management options by virtue of the superb neutronics and safety features of these heavy liquid metal (HLM) coolants. One option is once-for-life cores having relatively low power density. These cores are fueled in the factory; there is no refueling or fuel shuffling on site. A second option is very long-life cores being made of a fissioning zone and a natural uranium blanket zone. The fissioning zone very slowly drifts toward the blanket. A third option is multirecycling of light water reactor (LWR) discharged fuel without partitioning of transuranics (TRUs) in fuel-self-sustaining reactors. LWR spent fuel could provide the initial fuel loading after extracting fission products and {approx}90% of its uranium. The makeup fuel is natural or depleted uranium. A fourth option is the high-burnup once-through fuel cycle using natural or depleted uranium feed. The initial fuel loading of this reactor is a mixture of enriched and natural uranium. The natural uranium utilization is 10 to 20 times higher than that of a once-through LWR. A fifth option is transmutation of TRUs from LWRs using critical HLM-cooled reactors; such reactors could be designed to have the same high actinide burning capability of accelerator-driven systems and have comparable safety, but at a substantially lower cost. These novel reactor designs and fuel management options are hereby reviewed.

  1. Start-up fuel and power flattening of sodium-cooled candle core

    SciTech Connect

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-07-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

  2. PBF Reactor Building (PER620) basement. Camera facing north. Cooling air ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620) basement. Camera facing north. Cooling air compressor for control rods; inner cooler and after cooler; associated piping. Photographer: John Capek. Date: August 21, 1970. INEEL negative no. 70-3493 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  3. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  4. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  5. Long-term environmental trends: Selection of sampling locations in a reactor-aquatic cooling system

    SciTech Connect

    Revsin, B.K.; Watson, J.E. Jr. )

    1993-02-01

    The study objective was to determine whether environmental radionuclide accumulations were occurring in an aquatic system with a 13-y history of supplying a power plant with reactor-cooling water as well as receiving plant discharge. The aquatic system consisted of the following: (1) a reactor-cooling lake; (2) a secondary lake approximately 8 km downstream; and (3) a small stream that interfaced with the two lakes. Gamma-emitting radionuclides were identified and quantified in samples of benthic sediments obtained from representative areas of the aquatic system. This study demonstrated that in a reactor-aquatic cooling system, the component of the aquatic system most likely to experience radionuclide accumulation will not necessarily be the reactor-cooling lake, but will be that component of the aquatic system whose benthic sediments contain the highest concentrations of organic matter. Further, it was shown that the quantity of oxidizable organic matter present in a sediment is a good predictor or marker for potential sites of radionuclide accumulation (i.e., 60Co and 137Cs).

  6. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  7. Failed fuel identification techniques for liquid-metal cooled reactors

    SciTech Connect

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

    1995-06-01

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

  8. SEPARATING HAFNIUM FROM ZIRCONIUM

    DOEpatents

    Lister, B.A.J.; Duncan, J.F.; Hutcheon, J.M.

    1956-08-21

    Substantially complete separation of zirconium from hafnium may be obtained by elution of ion exchange material, on which compounds of the elements are adsorbed, with an approximately normal solution of sulfuric acid. Preferably the acid concentration is between 0.8 N amd 1.2 N, amd should not exceed 1.5 N;. Increasing the concentration of sulfate ion in the eluting solution by addition of a soluble sulfate, such as sodium sulfate, has been found to be advantageous. The preferred ion exchange materials are sulfonated polystyrene resins such as Dowex 50,'' and are preferably arranged in a column through which the solutions are passed.

  9. Control of reactor coolant flow path during reactor decay heat removal

    DOEpatents

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  10. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  11. Cooling by conversion of para to ortho-hydrogen

    NASA Technical Reports Server (NTRS)

    Sherman, A. (Inventor)

    1983-01-01

    The cooling capacity of a solid hydrogen cooling system is significantly increased by exposing vapor created during evaporation of a solid hydrogen mass to a catalyst and thereby accelerating the endothermic para-to-ortho transition of the vapor to equilibrium hydrogen. Catalyst such as nickel, copper, iron or metal hydride gels of films in a low pressure drop catalytic reactor are suitable for accelerating the endothermic para-to-ortho conversion.

  12. A thermodynamic approach for advanced fuels of gas-cooled reactors

    NASA Astrophysics Data System (ADS)

    Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  13. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-01-01

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  14. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-05-05

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  15. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  16. A combined gas cooled nuclear reactor and fuel cell cycle

    NASA Astrophysics Data System (ADS)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  17. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

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

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

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

  19. Conceptual design study of JSFR reactor building

    SciTech Connect

    Yamamoto, T.; Katoh, A.; Chikazawa, Y.; Ohya, T.; Iwasaki, M.; Hara, H.; Akiyama, Y.

    2012-07-01

    Japan Sodium-cooled Fast Reactor (JSFR) is planning to adopt the new concepts of reactor building. One is that the steel plate reinforced concrete is adopted for containment vessel and reactor building. The other is the advanced seismic isolation system. This paper describes the detail of new concepts for JSFR reactor building and engineering evaluation of the new concepts. (authors)

  20. ODS Ferritic/martensitic alloys for Sodium Fast Reactor fuel pin cladding

    NASA Astrophysics Data System (ADS)

    Dubuisson, Philippe; Carlan, Yann de; Garat, Véronique; Blat, Martine

    2012-09-01

    The development of ODS materials for the cladding for Sodium Fast Reactors is a key issue to achieve the objectives required for the GEN IV reactors. CEA, AREVA and EDF have launched in 2007 an important program to determine the optimal fabrication parameters, and to measure and understand the microstructure and properties before, under and after irradiation of such cladding materials. The aim of this paper is to present the French program and the major results obtained recently at CEA on Fe-9/14/18Cr1WTiY2O3 ferritic/martensitic ODS materials. The first step of the program was to consolidate Fe-9/14/18Cr ODS materials as plates and bars to study the microstructure and the mechanical properties of the new alloys. The second step consists in producing tubes at a geometry representative of the cladding of new Sodium Fast Reactors. The optimization of the fabrication route at the laboratory scale is conducted and different tubes were produced. Their microstructure depends on the martensitic (Fe-9Cr) or ferritic (Fe-14Cr) structure. To join the plug to the tube, the reference process is the welding resistance. A specific approach is developed to model the process and support the development of the welds performed within the "SOPRANO" facility. The development at CEA of Fe-9/14/18Cr new ODS materials for the cladding for GENIV Sodium Fast Reactors is in progress. The first microstructural and mechanical characterizations are very encouraging and the full assessment and qualification of this new alloys and products will pass through the irradiation of specimens, tubes, fuel pins and subassemblies up to high doses.